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Office of Scientific and Technical Information
ITER: The International Thermonuclear Experimental Reactor and the nuclear weapons proliferation implications of thermonuclear-fusion energy
E-Print Network (OSTI)
This paper contains two parts: (I) A list of "points" highlighting the strategic-political and military-technical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (II) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showing that while full access to the physics of thermonuclear weapons is the main implication of ICF, full access to large-scale tritium technology is the main proliferation impact of MCF. The conclusion of the paper is that siting ITER in a country such as Japan, which already has a large separated-plutonium stockpile, and an ambitious laser-driven ICF program (comparable in size and quality to those of the United States or France) will considerably increase its latent (or virtual) nuclear weapons proliferation status, and fo...
Gsponer, A; Gsponer, Andre; Hurni, Jean-Pierre
ITER: The International Thermonuclear Experimental Reactor and the Nuclear Weapons Proliferation Implications of Thermonuclear-Fusion Energy Systems
This paper contains two parts: (I) A list of �points � highlighting the strategic-political and militarytechnical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (II) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showing that while full access to the physics of thermonuclear weapons is the main implication of ICF, full access to large-scale tritium technology is the main proliferation impact of MCF. The conclusion of the paper is that siting ITER in a country such as Japan, which already has a large separated-plutonium stockpile, and an ambitious laser-driven ICF program (comparable in size and quality to those of the United States or France) will considerably increase its latent (or virtual) nuclear weapons proliferation status, and foster further nuclear proliferation throughout the world. The safety and environmental problems related to the operation of largescale fusion facilities such as ITER (which contain massive amounts of hazardous and/or radioactive materials such as tritium, lithium, and beryllium, as well as neutron-activated structural materials) are not addressed in this paper.
Andr� Gsponer; Jean-pierre Hurni
Structural characteristics of proposed ITER (International Thermonuclear Experimental Reactor) TF (toroidal field) coil conductor
SciTech Connect
This paper analyzes the effect of transverse loading on a cable-in-conduit conductor which has been proposed for the toroidal field coils of the International Thermonuclear Experimental Reactor. The primary components of this conductor are a loose cable of superconducting wires, a thin-wall tube for helium containment, and a U-shaped structural channel. A method is given where the geometry of this conductor can be optimized for a given set of operating conditions. It is shown, using finite-element modeling, that the structural channel is effective in supporting loads due to transverse forces and internal pressure. In addition, it is shown that the superconducting cable is effectively shielded from external transverse loads that might otherwise degrade its current carrying capacity. 10 refs., 10 figs., 3 tabs.
Gibson, C.R.; Miller, J.R.
A high-recycle divertor for ITER (International Thermonuclear Experimental Reactor)
A coupled one-dimensional (axial/radial) edge-plasma model (SOLAR) has been used to investigate tradeoffs between collector-plate and edge-plasma conditions in a doublenull, open, high-recycle divertor (HRD) for a preliminary International Thermonuclear Experimental Reactor (ITER) design. A steady-state HRD produces in attractive high-density edge plasma (5 /times/ 10/sup 19/m/sup /minus/3/) with sufficiently low plasma temperature (10-20eV) at a tungsten plat that the sheath-accelerated ions are below sputtering threshold energies. Manageable plate heat fluxes (3-6 MW/m/sup 2/) are achieved by positioning the plate poloidal cross section at a minimum angle of 15-30/degree/ with respect to flux surfaces. 6 refs., 9 figs.
Werley, K.A.; Bathke, C.G.
Review of the International Thermonuclear Experimental Reactor (ITER) detailed design report
Dr. Martha Krebs, Director, Office of Energy Research at the US Department of Energy (DOE), wrote to the Fusion Energy Sciences Advisory Committee (FESAC), in letters dated September 23 and November 6, 1996, requesting that FESAC review the International Thermonuclear Experimental Reactor (ITER) Detailed Design Report (DDR) and provide its view of the adequacy of the DDR as part of the basis for the United States decision to enter negotiations with the other interested Parties regarding the terms and conditions for an agreement for the construction, operations, exploitation and decommissioning of ITER. The letter from Dr. Krebs, referred to as the Charge Letter, provided context for the review and a set of questions of specific interest.
Requirements for US regulatory approval of the International Thermonuclear Experimental Reactor (ITER)
The International Thermonuclear Experimental Reactor (ITER) is the first fusion machine that will have sufficient decay heat and activation product inventory to pose potential nuclear safety concerns. As a result, nuclear safety and environmental issues will be much more important in the approval process for the design, siting, construction, and operation of ITER in the United States than previous fusion devices, such as the Tokamak Fusion Test Reactor. The purpose of this report is (a) to provide an overview of the regulatory approval process for a Department of Energy (DOE) nuclear facility; (b) to present the dose limits used by DOE to protect workers, the public, and the environment from the risks of exposure to radiation and hazardous materials; (c) to discuss some key nuclear safety-related issues that must be addressed early in the Engineering Design Activities (EDA) to obtain regulatory approval; and (d) to provide general guidelines to the ITER Joint Central Team (JCT) concerning the development of a regulatory framework for the ITER project.
Petti, D.A.; Haire, J.C.
Thermonuclear Reflect AB-Reactor
The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.
Alexander Bolonkin
International Thermonuclear Experimental Reactor U.S. Home Team Quality Assurance Plan
Science Conference Proceedings (OSTI)
The International Thermonuclear Experimental Reactor (ITER) project is unique in that the work is divided among an international Joint Central Team and four Home Teams, with the overall responsibility for the quality of activities performed during the project residing with the ITER Director. The ultimate responsibility for the adequacy of work performed on tasks assigned to the U.S. Home Team resides with the U.S. Home Team Leader and the U.S. Department of Energy Office of Fusion Energy (DOE-OFE). This document constitutes the quality assurance plan for the ITER U.S. Home Team. This plan describes the controls exercised by U.S. Home Team management and the Performing Institutions to ensure the quality of tasks performed and the data developed for the Engineering Design Activities assigned to the U.S. Home Team and, in particular, the Research and Development Large Projects (7). This plan addresses the DOE quality assurance requirements of 10 CFR 830.120, "Quality Assurance." The plan also describes U.S. Home Team quality commitments to the ITER Quality Assurance Program. The ITER Quality Assurance Program is based on the principles described in the International Atomic Energy Agency Standard No. 50-C-QA, "Quality Assurance for Safety in Nuclear Power Plants and Other Nuclear Facilities." Each commitment is supported with preferred implementation methodology that will be used in evaluating the task quality plans to be submitted by the Performing Institutions. The implementing provisions of the program are based on guidance provided in American National Standards Institute/American Society of Mechanical Engineers NQA-1 1994, "Quality Assurance." The individual Performing Institutions will implement the appropriate quality program provisions through their own established quality plans that have been reviewed and found to comply with U.S. Home Team quality assurance plan commitments to the ITER Quality Assurance Program. The extent of quality program provisions applied to any specific task is proportional to, and appropriate for, the safety and/or project success significance of the task, as determined by the cognizant Technical Manager and the U.S. Home Team Quality Coordinator. In general, the research and development activities will have only those controls appropriate to ensure the quality of the manufacturing activity and validate the resultant data.
Sowder, W. K.
03/01/2006 09:51 AMLoading "People's Daily Online --Chinese experimental thermonuclear reactor on discharge test in July" Page 1 of 1http://english.people.com.cn/200603/01/print20060301_247035.html
the commercialized nuclear reactors in the world were designed for fission, a process contrary to the ITER's fusion03/01/2006 09:51 AMLoading "People's Daily Online -- Chinese experimental thermonuclear reactor experimental thermonuclear reactor on discharge test in July China's new generation experimental Tokamak fusion
Controlled thermonuclear fusion reactors
Controlled production of energy by fusion of light nuclei has been the goal of a large portion of the physics community since the 1950's. In order for a fusion reaction to take place, the fuel must be heated to a temperature of 100 million degrees Celsius. At this temperature, matter can exist only in the form of an almost fully ionized plasma. In order for the reaction to produce net power, the product of the density and energy confinement time must exceed a minimum value of 10/sup 20/ sec m/sup -3/, the so-called Lawson criterion. Basically, two approaches are being taken to meet this criterion: inertial confinement and magnetic confinement. Inertial confinement is the basis of the laser fusion approach; a fuel pellet is imploded by intense laser beams from all sides and ignites. Magnetic confinement devices, which exist in a variety of geometries, rely upon electromagnetic forces on the charged particles of the plasma to keep the hot plasma from expanding. Of these devices, the most encouraging results have been achieved with a class of devices known as tokamaks. Recent successes with these devices have given plasma physicists confidence that scientific feasibility will be demonstrated in the next generation of tokamaks; however, an even larger effort will be required to make fusion power commercially feasible. As a result, emphasis in the controlled thermonuclear research program is beginning to shift from plasma physics to a new branch of nuclear engineering which can be called fusion engineering, in which instrumentation and control engineers will play a major role. Among the new problem areas they will deal with are plasma diagnostics and superconducting coil instrumentation.
Walstrom, P.L.
Physics problems of thermonuclear reactors
From topical conference on energy; Philadelphia, Pennsylvania, USA (Nov 1973). A problem common to all controlled fusion reactors is that of the burning deuterium --tritium fuel under conditions of plasma confinement which approach the ideal limit as nearly as possible. After ignitlon, the balance between alpha- particle energy deposition and plasma losses (radiation plus thermal and particle diffusion) determines the stability or instability of the burn in toroidal systems. Tokamak systems are described both with unstable, injection-regulated burn cycles and stabilized steady-state burn conditions. In the theta-pinch reactor an unstable burn occurs, some, what regulated by high-beta plasma expansion, which is quenched by a programmed plasma decompression. The plasma expansion during the constant-pressure burn provides direct'' conversion of plasma thermonuclear heat to electrical output, in addition to the electrical power derived from the neutron energy through conventional thermal conversion equipment. The open-ended mirror reactor is characterized by a direct conversion system for recovering end-loss plasma energy and converting it to electrical energy for reinjection into the plasma. This allows a favorable reactor energy balance and an amplification factor Q (= thermonuclear energy output/injected plasma energy) which is compatible with classical collisional losses. For the three reactor types considered the ramifications of burn and confinement conditions for reactor configuration, energy balance, economy, fuel handling, materials problems, and environmentalradiological factors are considered. (auth)
Ribe, F.L.
The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical pr...
Bolonkin, Alexander
Evaluation of graphite/steam interactions for ITER (International Thermonuclear Experimental Reactor)
DOE Green Energy (OSTI)
In this report we present the results of an experimental/analytical study designed to determine the quantity of hydrogen generated during a coolant inleakage accident in ITER. This hydrogen could represent a potential explosive hazard, provided the proper conditions exist, causing machine damage and release of radioactive material. We have measured graphite/steam reaction rates for several graphites and carbon-based composites at temperatures between 1000 C and 1700 C. The effects of steam flow rate, and partial pressure were also examined. The measured reaction rates correlated well with two Arrhenius type relationships. We have used the relationships for GraphNOL N3M in a thermal model to determine that for ITER the quantity of hydrogen produced would range between 5 and 35 kg, depending upon how the graphite tiles are attached to the first wall. While 5 kg is not a significant concern, 35 kg presents an explosive hazard. 20 refs., 14 figs., 1 tab.
Smolik, G.R.; Merrill, B.J.; Piet, S.J.; Holland, D.F.
Micro -Thermonuclear AB-Reactors for Aerospace
The author offers several innovations that he first suggested publicly early in 1983 for the AB multi-reflex engine, space propulsion, getting energy from plasma, etc. (see: A. Bolonkin, Non-Rocket Space Launch and Flight, Elsevier, London, 2006, Chapters 12, 3A). It is the micro-thermonuclear AB-Reactors. That is new micro-thermonuclear reactor with very small fuel pellet that uses plasma confinement generated by multi-reflection of laser beam or its own magnetic field. The Lawson criterion increases by hundreds of times. The author also suggests a new method of heating the power-making fuel pellet by outer electric current as well as new direct method of transformation of ion kinetic energy into harvestable electricity. These offered innovations dramatically decrease the size, weight and cost of thermonuclear reactor, installation, propulsion system and electric generator. Non-industrial countries can produce these researches and constructions. Currently, the author is researching the efficiency of these innovations for two types of the micro-thermonuclear reactors: multi-reflection reactor (ICF) and self-magnetic reactor (MCF).
New AB-Thermonuclear Reactor for Aerospace
There are two main methods of nulcear fusion: inertial confinement fusion (ICF) and magnetic confinement fusion (MCF). Existing thermonuclear reactors are very complex, expensive, large, and heavy. They cannot achieve the Lawson creterion. The author offers an innovation. ICF has on the inside surface of the shell-shaped combustion chamber a covering of small Prism Reflectors (PR) and plasma reflector. These prism reflectors have a noteworthy advantage, in comparison with conventional mirror and especially with conventional shell: they multi-reflect the heat and laser radiation exactly back into collision with the fuel target capsule (pellet). The plasma reflector reflects the Bremsstrahlung radiation. The offered innovation decreases radiation losses, creates significant radiation pressure and increases the reaction time. The Lawson criterion increases by hundreds of times. The size, cost, and weight of a typical installation will decrease by tens of times. The author is researching the efficiency of these innovations. Keywords: Thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, aerospace thermonuclear engine. This work is presented as paper AIAA-2006-7225 to Space-2006 Conference, 19-21 September, 2006, San Jose, CA, USA.
There are two main methods of nulcear fusion: inertial confinement fusion (ICF) and magnetic confinement fusion (MCF). Existing thermonuclear reactors are very complex, expensive, large, and heavy. They cannot achieve the Lawson creterion. The author offers an innovation. ICF has on the inside surface of the shell-shaped combustion chamber a covering of small Prism Reflectors (PR) and plasma reflector. These prism reflectors have a noteworthy advantage, in comparison with conventional mirror and especially with conventional shell: they multi-reflect the heat and laser radiation exactly back into collision with the fuel target capsule (pellet). The plasma reflector reflects the Bremsstrahlung radiation. The offered innovation decreases radiation losses, creates significant radiation pressure and increases the reaction time. The Lawson criterion increases by hundreds of times. The size, cost, and weight of a typical installation will decrease by tens of times. The author is researching the efficiency of these i...
Application of controlled thermonuclear reactor fusion energy for food production
Food and energy shortages in many parts of the world in the past two years raise an immediate need for the evaluation of energy input in food production. The present paper investigates systematically (1) the energy requirement for food production, and (2) the provision of controlled thermonuclear fusion energy for major energy intensive sectors of food manufacturing. Among all the items of energy input to the ''food industry,'' fertilizers, water for irrigation, food processing industries, such as beet sugar refinery and dough making and single cell protein manufacturing, have been chosen for study in detail. A controlled thermonuclear power reactor was used to provide electrical and thermal energy for all these processes. Conceptual design of the application of controlled thermonuclear power, water and air for methanol and ammonia synthesis and single cell protein production is presented. Economic analysis shows that these processes can be competitive. (auth)
Dang, V.D.; Steinberg, M.
Iran cannot have claimed to be building a thermonuclear reactor -To build a thermonuclear reactor by its own efforts is a task absolutely beyond Iran's
Iran cannot have claimed to be building a thermonuclear reactor - expert To build a thermonuclear reactor by its own efforts is a task absolutely beyond Iran's strength, and reports that Iran has made such claims must be a misunderstanding, a nuclear specialist said._Georgy Yeliseyev, chief specialist
Stars as thermonuclear reactors: their fuels and ashes
Atomic nuclei are transformed into each other in the cosmos by nuclear reactions inside stars: -- the process of nucleosynthesis. The basic concepts of determining nuclear reaction rates inside stars and how they manage to burn their fuel so slowly most of the time are discussed. Thermonuclear reactions involving protons in the hydrostatic burning of hydrogen in stars are discussed first. This is followed by triple alpha reactions in the helium burning stage and the issues of survival of carbon and oxygen in red giant stars connected with nuclear structure of oxygen and neon. Advanced stages of nuclear burning in quiescent reactions involving carbon, neon, oxygen and silicon are discussed. The role of neutron induced reactions in nucleosynthesis beyond iron is discussed briefly, as also the experimental detection of neutrinos from SN 1987A which confirmed broadly the ideas concerning gravitational collapse leading to a supernova.
A. Ray
in a joint JCT/Home Team �?Concept Improvement Task Force�? constituted in April 1999, following the guiding will address confinement, stability, exhaust of helium ash, and impurity control in plasmas dominated by alpha/semi-detached divertor, ELMy H-mode, etc.); - particle control (fuel/ash/impurity/fuelling/pumping); - steady
Finite element modeling and experimental study of brittle fracture in tempered martensitic steels for thermonuclear fusion applications.
??In this work we have studied brittle fracture in high-chromium reduced activation tempered martensitic steels foreseen as structural materials for thermonuclear fusion reactors. Developing the� (more)
Mueller, Pablo Federico
Astrophysical thermonuclear functions
Stars are gravitationally stabilized fusion reactors changing their chemical composition while transforming light atomic nuclei into heavy ones. The atomic nuclei are supposed to be in thermal equilibrium with the ambient plasma. The majority of reactions among nuclei leading to a nuclear transformation are inhibited by the necessity for the charged participants to tunnel through their mutual Coulomb barrier. As theoretical knowledge and experimental verification of nuclear cross sections increases it becomes possible to refine analytic representations for nuclear reaction rates. Over the years various approaches have been made to derive closed-form representations of thermonuclear reaction rates (Critchfield 1972, Haubold and John 1978, Haubold, Mathai and Anderson 1987). They show that the reaction rate contains the astrophysical cross section factor and its derivatives which has to be determined experimentally, and an integral part of the thermonuclear reaction rate independent from experimental results which can be treated by closed-form representation techniques in terms of generalized hypergeometric functions. In this paper mathematical/statistical techniques for deriving closed-form representations of thermonuclear functions will be summarized and numerical results for them will be given. The separation of thermonuclear functions from thermonuclear reaction rates is our preferred result. The purpose of the paper is also to compare numerical results for approximate and closed-form representations of thermonuclear functions. This paper completes the work of Haubold, Mathai, and Anderson (1987).
William J. Anderson; Hans J. Haubold; Arak Mathai Mathai
Thermonuclear inverse magnetic pumping power cycle for stellarator reactor
DOE Patents (OSTI)
The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor.
Ho, Darwin D. (Pleasanton, CA); Kulsrud, Russell M. (Princeton, NJ)
The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor. 9 figs., 4 tabs.
Ho, D.D.M.; Kulsrud, R.M.
Lower hybrid current drive at plasma densities required for thermonuclear reactors
Driving current in high-density plasmas is essential for the progress of thermonuclear fusion energy research based on the tokamak concept. The lower hybrid current drive (LHCD) effect, is potentially the most suitable tool for driving current at large plasma radii, consistent with the needs of ITER steady state scenario. Unfortunately, experiments at reactor grade high plasma densities with kinetic profiles approaching those required for ITER, have shown problems in penetration of the LH power into the core plasma. These plasmas represent a basic reference for designing possible methods useful for assessing the LHCD concept in ITER. On the basis of the phenomenology observed during LHCD experiments carried out in different machines, and model of the spectral broadening effect due to parametric instability, an interpretation and possible solution of the related important problem is presented.
Cesario, R.; Cardinali, A.; Castaldo, C.; Tuccillo, A. A.; Amicucci, L. [Associazione EURATOM/ENEA sulla Fusione, Centro Ricerche Frascati, 00044 Frascati (Italy)
Energy and particle confinement in ITER (International Thermonuclear Experimental Reactor)
ITER will require a level of energy ({tau}{sub E}) and particle ({tau}{sub p}) confinement sufficient for reaching ignition and extended burn of DT plasmas, with steady-state, high-Q (Q = P{sub fusion}/P{sub aux} = 5P{sub {alpha}}/P{sub aux}) operation as an ultimate goal. This translates into a required confinement capability of T{sub i}(0)n{sub DT}(0){tau}{sub E} {approximately}4--8 {times} 10{sup 21} keV{center dot}s/m{sup 3} for Q {approximately}5--{infinity} (where T{sub i}(0) is the central ion temperature, n{sub DT}(0) is the central DT fuel ion density, {tau}{sub E} is the global energy confinement time). In addition, {tau}{sub p} {sup He}/{tau}{sub E} < 10 is required to ensure that the thermal {alpha}-particle (He) accumulation (n{sub He}/n{sub e}) is less than 10% (where {tau}{sub p}{sup He} is the global thermal alpha particle containment time). Furthermore, the tritium fuel burnup fraction under nominal burn conditions must be greater than 1.5%. While the ITER Tn{tau}{sub E} requirement is about ten times the highest value achieved in JET (3), the extrapolation in plasma physics parameters is less. If {beta} (ratio of plasma kinetic pressure to magnetic pressure), {nu}* (collisionality), and {rho}/a (ratio of the ion gyroradius to the plasma minor radius) are taken as the relevant dimensionless parameters to characterize the plasma, the extrapolation from present JET performance to ITER is a factor {le}2. The ITER concept is based on the expectation that H-mode confinement can be achieved for long pulse. This paper covers the energy and particle confinement issues for ITER. 15 refs.
Uckan, N.A. (Oak Ridge National Lab., TN (USA)); Yushmanov, P.N.; Mukhovatov, V.S.; Parail, V.V.; Putvinskii, S. (Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Moscow (USSR). Inst. Atomnoj Ehnergii); Takizuka, T.; Tsunematsu, T.; Yoshino, R. (Japan Atomic Energy Research Inst., Tokyo (Japan)); Borrass, K.; Engelmann, F.; Kardaun, O.; Lackner, K. (Max-Planck-Institut fuer Plasmaphysik,
International Thermonuclear Experimental Reactor (ITER) neutral beam design
This report discusses the following topics on ITER neutral beam design: ion dump; neutralizer and module gas flow analysis; vacuum system; cryogenic system; maintainability; power distribution; and system cost.
Myers, T.J.; Brook, J.W.; Spampinato, P.T.; Mueller, J.P.; Luzzi, T.E.; Sedgley, D.W. (Grumman Corp., Bethpage, NY (USA). Space Systems Div.)
ITER: Image Collections from the International Thermonuclear Experimental Reactor (ITER)
DOE Data Explorer (OSTI)
ITER is a large-scale scientific experiment that aims to demonstrate that it is possible to produce commercial energy from fusion. The Q in the formula symbolizes the ratio of fusion power to input power. Q = 10 represents the scientific goal of the ITER project: to deliver ten times the power it consumes. From 50 MW of input power, the ITER machine is designed to produce 500 MW of fusion power - the first of all fusion experiments to produce net energy. During its operational lifetime, ITER will test key technologies necessary for the next step: the demonstration fusion power plant that will prove that it is possible to capture fusion energy for commercial use. The science going on at ITER - and all around the world in support of ITER - will benefit all of mankind. "We firmly believe that to harness fusion energy is the only way to reconcile huge conflicting demands which will confront humanity sooner or later. The issue at stake is how to reconcile the imperative, constantly growing demand of the majority of the world's population to raise their standard of living with the enormous environmental hazards resulting from the present energy supply. In our opinion, the use of fusion energy is a "must" if we want to be serious about embarking on sustainable development for future generations." Director-General Osamu Motojima, Opening address, Monaco International ITER Fusion Energy Days (MIIFED), 23 November 2010 [copied from http://www.iter.org/proj/itermission]. More than 900 images, organized into categories, are available as of 2013.
One-dimensional thermonuclear burn computations for the Reversed-Field Pinch Reactor (RFPR)
Conceptual fusion reactor designs of the Reversed-Field Pinch Reactor (RFPR) have been based on profile-averaged zero-dimensional (point) plasma models. The plasma response/performance that has been predicted by the point plasma model is re-examined by a comprehensive one-dimensional (radial) burn code (RFPBRN) that has been developed and parametrically evaluated for the RFPR. The RFPR plasma parameters have been optimized and effects of turbulent transport and stability have been studied.
Nebel, R.A.; Miley, G.H.; Moses, R.W.
Concept of a thermonuclear reactor based on gravity retention of high-temperature plasma
In the present paper the realization of the obtained results in relation to the dense high- temperature plasma of multivalent ions including experimental data interpretation is discussed.
S. I. Fisenko; I. S. Fisenko
Method and system to directly produce electrical power within the lithium blanket region of a magnetically confined, deuterium-tritium (DT) fueled, thermonuclear fusion reactor
A method for integrating liquid metal magnetohydrodynamic power generation with fusion blanket technology to produce electrical power from a thermonuclear fusion reactor located within a confining magnetic field and within a toroidal structure. A hot liquid metal flows from a liquid metal blanket region into a pump duct of an electromagnetic pump which moves the liquid metal to a mixer where a gas of predetermined pressure is mixed with the pressurized liquid metal to form a Froth mixture. Electrical power is generated by flowing the Froth mixture between electrodes in a generator duct. When the Froth mixture exits the generator the gas is separated from the liquid metal and both are recycled.
Woolley, Robert D. (Belle Mead, NJ)
Method and System to Directly Produce Electrical Power within the Lithium Blanket Region of a Magnetically Confined, Deuterium-Tritium (DT) Fueled, Thermonuclear Fusion Reactor
Woolley, Robert D.
Experimental Breeder Reactor I Preservation Plan
Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.
Julie Braun
International thermonuclear experimental reactor-like extended Solovev equilibria with parallel flow
A Solovev-like equilibrium solution is extended to plasmas with incompressible flow parallel to the magnetic field. ITER-like configurations are constructed for Alfven Mach functions peaked either on or off the magnetic axis. The linear stability of the equilibrium is also examined by applying a sufficient condition.
Throumoulopoulos, G. N. [Section of Theoretical Physics, University of Ioannina, Association Euratom-Hellenic Republic, GR 451 10 Ioannina (Greece); Tasso, H. [Max-Planck-Institut fuer Plasmaphysik, Euratom Association, D-85748 Garching (Germany)
Comparison of zero-dimensional and one-dimensional thermonuclear burn computations for the reversed-field pinch reactor (RFPR)
Conceptual fusion reactor designs of the Reversed-Field Pinch Reactor (RFPR) have been based on profile-averaged zero-dimensional (point) plasma models. The plasma response/performance that has been predicted by the point plasma model is re-examined by a comprehensive one-dimensional (radial) burn code that has been developed and parametrically evaluated for the RFPR. Agreement is good between the zero-dimensional and one-dimensional models, giving more confidence in the RFPR design point reported previously from the zero-dimensional analysis.
Nebel, R.A.; Hagenson, R.L.; Moses, R.W.; Krakowski, R.A.
Reactor /Commercial Reactor Engineering Design Expending Performance Phase High Working Rate CDA EDA are condensed under the Earth during a few hundred millions of year. This precious gift will be expired within.) Disposal cost for radioactive materials is one of important element for commercial use of fusion. Figure 9
Thermal instability of thermonuclear plasma in a mirror field
In this paper, the thermal stability of a thermonuclear plasma in a mirror reactor is obtained by a simple model. The effect of the loss of thermonuclear alpha particles due to collisional pitch-angel scattering into loss cones is included in this analysis. The effect of the collisional loss is significant, and it has a stabilizing effect on the thermal instability.
Mizuno, N. (Nihon Univ., Tokyo (Japan). Coll. of Science and Engineering)
Simplest AB-Thermonuclear Space Propulsion and Electric Generator
The author applies, develops and researches mini-sized Micro- AB Thermonuclear Reactors for space propulsion and space power systems. These small engines directly convert the high speed charged particles produced in the thermonuclear reactor into vehicle thrust or vehicle electricity with maximum efficiency. The simplest AB-thermonuclear propulsion offered allows spaceships to reach speeds of 20,000 50,000 km/s (1/6 of light speed) for fuel ratio 0.1 and produces a huge amount of useful electric energy. Offered propulsion system permits flight to any planet of our Solar system in short time and to the nearest non-Sun stars by E-being or intellectual robots during a single human life period. Key words: AB-propulsion, thermonuclear propulsion, space propulsion, thermonuclear power system.
Bolonkin, A
Controlled thermonuclear fusion research in Europe -- Competence in advanced physics and technologies
Development of Fusion power is being pursued in all major industrial countries. The European Union, together with countries associated to the EURATOM-Framework Program undertakes an integrated RTD program for the development of magnetic fusion. The Key Action Controlled Thermonuclear Fusion has the objectives to develop the capacity to construct and operated a Next Step device for which the design is being pursued in international collaboration (ITER EDA, International Thermonuclear Experimental Reactor Engineering Design Activities); to undertake structured activities for concept improvements for a fusion power station; to develop technologies needed in the longer term for a prototype fusion reactor. Work on the socio-economic aspects of fusion and a keep in touch activity coordinating national civil research activities in inertial confinement fusion complement the program.
Bruhns, H.
>> Home >> World News Michael Richardson: Win for France over fusion reactor a big blow to Japan
, should be the site for an experimental nuclear fusion reactor. But the choice was a loss for big scienceWorld News >> Home >> World News Michael Richardson: Win for France over fusion reactor a big blow, the International Thermonuclear Experimental Reactor. Due largely to China's opposition, ITER will be constructed
Alerting Humanitarians to Emergencies PARIS, Jan 12 (Reuters) -Europe could build a disputed experimental nuclear
experimental nuclear fusion reactor on its own if the multinational joint venture promoting it decided to give, as opposed to today's nuclear reactors and weapons, which produce energy by splitting atoms apart. Fifty International Thermonuclear Experimental Reactor (ITER), will not give up its bid to host the programme, Prime
A preliminary evaluation of the hazards associated with a 20-Mw boiling reactor for the purpose of determining site requirements is presented. The Boiling Experimental Reactor design, safety features, and performance are given and the surroundings of the site at Argonne National Laboratory are described. (T.R.H.)
West, J.M.; Anderson, C.A.; Dietrich, J.R.; Harrer, J.M.; Jameson, A.S.; Untermyer, S.
President Truman Orders Development of Thermonuclear Weapon ...
National Nuclear Security Administration (NNSA)
Weapon President Truman Orders Development of Thermonuclear Weapon January 31, 1950 Washington, DC President Truman Orders Development of Thermonuclear Weapon President...
Decontamination and decommissioning experience at Experimental Breeder Reactor No. 1
A description is presented of the experience obtained from decontamination and decommissioning of the Experimental Breeder Reactor No. 1 located at the Idaho Nuclear Engineering Laboratory, Scottsville, Idaho. Included are the planning, preparation, and operations associated with the removal and processing of radioactively contaminated sodium-potassium eutectic alloy (NaK) and decontamination and decommissioning of the facility. (auth)
Kendall, E.W.
Revised design for the Tokamak experimental power reactor
A new, preliminary design has been identified for the tokamak experimental power reactor (EPR). The revised EPR design is simpler, more compact, less expensive and has somewhat better performance characteristics than the previous design, yet retains many of the previously developed design concepts. This report summarizes the principle features of the new EPR design, including performance and cost.
Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.
Experimental Requirements to Determine the Neutrino Mass Hierarchy Using Reactor Neutrinos
This paper presents experimental requirements to determine the neutrino mass hierarchy using reactor neutrinos. The detector shall be located at a baseline around 58 km from the reactor(s) to measure the energy spectrum of electron antineutrinos ($\\bar{\
Liang Zhan; Yifang Wang; Jun Cao; Liangjian Wen
EBR-2 (Experimental Breeder Reactor-2) test programs
The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.
Sackett, J.I.; Lehto, W.K.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA)); Planchon, H.P.; Lambert, J.D.B.; Hill, D.J. (Argonne National Lab., IL (USA))
LIUM OXIDE REACTOR. MARITIME GAS-COOLED The Experimental Beryllium Oxide Reactor, EBOR, will be constructed at the National Reactor Testing Station as the AEC portion of the joint Maritime Administration--AEC Maritime Gas Cooled Reactor Program. The ultimate goal of the Program is the development of nuclear power plants employing a helium cooled and beryllium oxide moderated reactor directly coupled to a closed cycle gas turbine. The objective is to obtain compact nuclear engines suitable for use either in a merchant ship propulsion system or an intermediate size central station power plant in the 20 to 100 Mw(e) size range. The EBOR is a l0 Mw(t) test of the basic fuel element and moderator designs. It is capable of being up-graded in power at a later date to a test of the nuclear reactor turbine concept. The objective of the experiment is outlined. The principal reactor components to be tested and the test facility are described. (auth)
Moore, W.C.
Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation
The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).
Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri
Plasma driving system for a Tokamak Experimental Power Reactor
The Tokamak Experimental Power Reactor is a device designed to operate at or near ignition with the potential for production of net power. It utilizes technology which either exists or can be achieved with a modest extrapolation from the present. The plasma driving systems, and their implications for energy storage and transfer technology are analyzed utilizing dynamic simulation of the plasma startup, burn and shutdown.
Mills, F.E.; Brooks, J.N.; Evans, K. Jr.; Kustom, R.L.; Stacey, W.M. Jr.; Wang, S.T.
Flash code: studying astrophysical thermonuclear flashes
The Center for Astrophysical Thermonuclear Flashes is constructing a new generation of codes designed to study runaway thermonuclear burning on the surface or in the interior of evolved compact stars.
Robert Rosner; Alan Calder; Jonathan Dursi; Bruce Fryxell; Donald Q. Lamb; Jens C. Niemeyer; Kevin Olson; Paul Ricker; Frank X. Timmes; James W. Truran; Henry Tufo; Yuan-Nan Young; Michael Zingale; Ewing Lusk; Rick Stevens
A Study of Reactor Neutrino Monitoring at Experimental Fast Reactor JOYO
We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3m from the JOYO reactor core of 140MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11\\pm1.24(stat.)\\pm0.46(syst.)events/day. Although the statistical significance of the measurement was not enough, the background in such a compact detector at the ground level was studied in detail and MC simulation was found to describe the data well. A study for improvement of the detector for future such experiments is also shown.
H. Furuta; Y. Fukuda; T. Hara; T. Haruna; N. Ishihara; M. Ishitsuka; C. Ito; M. Katsumata; T. Kawasaki; T. Konno; M. Kuze; J. Maeda; T. Matsubara; H. Miyata; Y. Nagasaka; K. Nitta; Y. Sakamoto; F. Suekane; T. Sumiyoshi; H. Tabata; M. Takamatsu; N. Tamura
EBR-2 (Experimental Breeder Reactor-2), IFR (Integral Fast Reactor) prototype testing programs
The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.
Lehto, W.K.; Sackett, J.I.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA). EBR-II Div. Argonne National Lab., IL (USA)); Planchon, H.P.; Lambert, J.D.B. (Argonne National Lab., IL (USA))
Measurement of neutron capture on $^{48}$Ca at thermal and thermonuclear energies
At the Karlsruhe pulsed 3.75\\,MV Van de Graaff accelerator the thermonuclear $^{48}$Ca(n,$\\gamma$)$^{49}$Ca(8.72\\,min) cross section was measured by the fast cyclic activation technique via the 3084.5\\,keV $\\gamma$-ray line of the $^{49}$Ca-decay. Samples of CaCO$_3$ enriched in $^{48}$Ca by 77.87\\,\\% were irradiated between two gold foils which served as capture standards. The capture cross-section was measured at the neutron energies 25, 151, 176, and 218\\,keV, respectively. Additionally, the thermal capture cross-section was measured at the reactor BR1 in Mol, Belgium, via the prompt and decay $\\gamma$-ray lines using the same target material. The $^{48}$Ca(n,$\\gamma$)$^{49}$Ca cross-section in the thermonuclear and thermal energy range has been calculated using the direct-capture model combined with folding potentials. The potential strengths are adjusted to the scattering length and the binding energies of the final states in $^{49}$Ca. The small coherent elastic cross section of $^{48}$Ca+n is explained through the nuclear Ramsauer effect. Spectroscopic factors of $^{49}$Ca have been extracted from the thermal capture cross-section with better accuracy than from a recent (d,p) experiment. Within the uncertainties both results are in agreement. The non-resonant thermal and thermonuclear experimental data for this reaction can be reproduced using the direct-capture model. A possible interference with a resonant contribution is discussed. The neutron spectroscopic factors of $^{49}$Ca determined from shell-model calculations are compared with the values extracted from the experimental cross sections for $^{48}$Ca(d,p)$^{49}$Ca and $^{48}$Ca(n,$\\gamma$)$^{49}$Ca.
H. Beer; C. Coceva; P. V. Sedyshev; Yu. P. Popov; H. Herndl; R. Hofinger; P. Mohr; H. Oberhummer
Experimental power reactor dc generator energy storage study
This study covers the use of dc generators for meeting the Experimental Power Reactor Ohmic Heating Energy Storage Requirements. The dc generators satisfy these requirements which are the same as defined in WFPS-TME-038 which covered the use of ac generators and homopolar generators. The costs of the latter two systems have been revised to eliminate first-of-a-kind factors. The cost figures for dc generators indicate a need to develop larger machines in order to take advantage of the economy-of-scale that the large ac machines have. Each of the systems has its own favorable salient features on which to base a system selection.
Heck, F.M.; Smeltzer, G.S.; Myers, E.H.; Kilgore, L.
Hydrogen Isotope Separation System for the Tokamak Experimental Power Reactor
An isotopic separation system for processing the fuel in the Tokamak Experimental Power Reactor is described. Two cryogenic distillation columns are used in sequence to recover 80% of the hydrogen from a fuel mixture originally containing equal parts of deuterium and tritium with a 1% hydrogen impurity. The hydrogen thus removed contains less than 1/2% tritium, which may be recovered in a separate system designed for that purpose. It is assumed that separation of the deuterium and the tritium is not required. A total tritium inventory of approximately 38,000 Ci (3.8 g) is projected.
Wilkes, W. R.
Experimental studies of tritium barrier concepts for fusion reactors
Ongoing experimental studies at ANL aimed at the development of methods to reduce tritium migration in fusion reactor systems currently include (1) work on the development of multilayered metal composites and impurity-coated refractory metals as barriers to tritium permeation in elevated temperature (greater than 300$sup 0$C) structures and (2) investigations of the kinetics of tritium trapping reactions in inert gas purge streams under conditions that emulate fusion reactor environments. Significant results obtained thus far are (1) demonstration of greater than 50-fold reductions in the hydrogen permeability of stainless steel structures by using stainless steel-clad composites containing an intermediate layer of a selected copper alloy and (2) verification that surface-oxide coatings lead to greater than 100-fold reductions in the hydrogen permeability of vanadium, but that severe oxygen penetration and embrittlement of the vanadium occur at temperatures in the range from 300 to 800$sup 0$C and under conditions of extremely low oxygen potential. Other considerations pertaining to the large-scale use of metal composites in fusion reactors are discussed, and progress in efforts to demonstrate the fabricability of metal composites is reviewed. Also presented are results of studies of the efficiencies of (1) CuO and CuO-MnO$sub 2$ beds in converting HT to HTO and (2) magnesium metal beds in converting HTO to HT. (auth)
Maroni, V.A.; Van Deventer, E.H.; Renner, T.A.; Pelto, R.H.; Wierdak, C.J.
Experimental investigation of creep behavior of reactor vessel lower head
The objective of the USNRC supported Lower Head Failure (LHF) Experiment Program at Sandia National Laboratories is to experimentally investigate and characterize the failure of the reactor pressure vessel (RPV) lower head due to the thermal and pressure loads of a severe accident. The experimental program is complemented by a modeling program focused on the development of a constitutive formulation for use in standard finite element structure mechanics codes. The problem is of importance because: lower head failure defines the initial conditions of all ex-vessel events; the inability of state-of-the-art models to simulate the result of the TMI-II accident (Stickler, et al. 1993); and TMI-II results suggest the possibility of in-vessel cooling, and creep deformation may be a precursor to water ingression leading to in-vessel cooling.
Chu, T.Y.; Pilch, M.; Bentz, J.H. [Sandia National Labs., Albuquerque, NM (United States); Behbahani, A. [NRC, Washington, DC (United States)
Shutdown and closure of the experimental breeder reactor - II.
The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m{sup 3} (86,000 gallons) of sodium and the secondary system contained 50 m{sub 3} (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and maintenance requirements during the interim period between deactivation and decommissioning. The plans also establish document archival of not only all the closure documents, but also the key plant documents (P and IDs, design bases, characterization data, etc.) in a convenient location to assure the appropriate knowledge base is available for decommissioning, which could occur decades in the future.
Michelbacher, J. A.; Baily, C. E.; Baird, D. K.; Henslee, S. P.; Knight, C. J.; Rosenberg, K. E.
Liquid metal reactor deactivation as applied to the experimental breeder reactor - II.
The Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-W) was shutdown in September, 1994. This sodium cooled reactor had been in service since 1964, and by the US Department of Energy (DOE) mandate, was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility (SPF) was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the SPF.
Earle, O. K.; Michelbacher, J. A.; Pfannenstiel, D. F.; Wells, P. B.
An elementary explanation of the stages which led to the controlled thermonuclear reactions at Harwell is given. (M H.R.)
Thomson, G.
First Thermonuclear Device Successfully Tested | National Nuclear...
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Experimental Breeder Reactor-II Primary Tank System Wash Water Workshop
Energy.gov (U.S. Department of Energy (DOE))
In 1994 Congress ordered the shutdown of the Experimental Breeder Reactor-II (EBR-II) and a closure project was initiated.
Fusion reactor materials semiannual progress report for the period ending March 31, 1990
This report mainly discusses topics on the physical effects of radiation on thermonuclear reactor materials. The areas discussed are: irradiation facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; fundamental mechanical behavior; radiation effects; mechanistic studies, theory and modeling; development of structural alloys; solid breeding materials; and ceramics. (FI)
Not Available
Decommissioning of Experimental Breeder Reactor - II Complex, Post Sodium Draining
The Experimental Breeder Reactor - II (EBR-II) was shutdown in September 1994 as mandated by the United States Department of Energy. This sodium-cooled reactor had been in service since 1964. The bulk sodium was drained from the primary and secondary systems and processed. Residual sodium remaining in the systems after draining was converted into sodium bicarbonate using humid carbon dioxide. This technique was tested at Argonne National Laboratory in Illinois under controlled conditions, then demonstrated on a larger scale by treating residual sodium within the EBR-II secondary cooling system, followed by the primary tank. This process, terminated in 2002, was used to place a layer of sodium bicarbonate over all exposed surfaces of sodium. Treatment of the remaining EBR-II sodium is governed by the Resource Conservation and Recovery Act (RCRA). The Idaho Department of Environmental Quality issued a RCRA Operating Permit in 2002, mandating that all hazardous materials be removed from EBR-II within a 10 year period, with the ability to extend the permit and treatment period for another 10 years. A preliminary plan has been formulated to remove the remaining sodium and NaK from the primary and secondary systems using moist carbon dioxide, steam and nitrogen, and a water flush. The moist carbon dioxide treatment was resumed in May 2004. As of August 2005, approximately 60% of the residual sodium within the EBR-II primary tank had been treated. This process will continue through the end of 2005, when it is forecast that the process will become increasingly ineffective. At that time, subsequent treatment processes will be planned and initiated. It should be noted that the processes and anticipated costs associated with these processes are preliminary. Detailed engineering has not been performed, and approval for these methods has not been obtained from the regulator or the sponsors.
J. A. (Bart) Michelbacher; S. Paul Henslee; Collin J. Knight; Steven R. sherman
Public comment sought on final end state of Experimental Breeder Reactor-II
NLE Websites -- All DOE Office Websites (Extended Search)
Media Contacts: Danielle Miller, 208-526-5709 Media Contacts: Danielle Miller, 208-526-5709 Joseph Campbell, CWI, 208-360-0142 Public comment sought on final end state of Experimental Breeder Reactor-II The U.S. Department of Energy (DOE) is seeking public comment on a range of alternatives for disposition of the landmark Experimental Breeder Reactor-II (EBR-II) building and reactor vessel at the Idaho Site's Materials and Fuels Complex. An Engineering Evaluation/Cost Analysis (EE/CA) document with four proposed alternatives for the final end state of the reactor facility and support structures is currently under evaluation by DOE, the U.S. Environmental Protection Agency, and Idaho's Department of Environmental Quality. Experimental Breeder Reactor-II containment dome The EBR-II was an innovative sodium-cooled reactor with an output of 62
Controlled fusion physics: experimental
A historical review is given of the experimental thermonuclear research program. The role of pinch devices, mirror machines, tokamak devices, and laser fusion is discussed. (MOW)
Post, R.F.
The objective of this study was to develop a predictive experimentally verified computational fluid dynamics (CFD) model for gas-liquid-solid flow. A three dimensional transient computer code for the coupled Navier-Stokes equations for each phase was developed and is appended in this report. The principal input into the model is the viscosity of the particulate phase which was determined from a measurement of the random kinetic energy of the 800 micron glass beads and a Brookfield viscometer. The details are presented in the attached paper titled ''CFD Simulation of Flow and Turbulence in a Slurry Bubble Column''. This phase of the work is in press in a referred journal (AIChE Journal, 2002) and was presented at the Fourth International Conference on Multiphase Flow (ICMF 2001) in New Orleans, May 27-June 1, 2001 (Paper No. 909). The computed time averaged particle velocities and concentrations agree with Particle Image Velocimetry (PIV) measurements of velocities and concentrations, obtained using a combination of gamma-ray and X-ray densitometers, in a slurry bubble column, operated in the bubbly-coalesced fluidization regime with continuous flow of water. Both the experiment and the simulation show a down-flow of particles in the center of the column and up-flow near the walls and nearly uniform particle concentration. Normal and shear Reynolds stresses were constructed from the computed instantaneous particle velocities. The PIV measurement and the simulation produced instantaneous particle velocities. The PIV measurement and the simulation produced similar nearly flat horizontal profiles of turbulent kinetic energy of particles. To better understand turbulence we studied fluidization in a liquid-solid bed. This work was also presented at the Fourth International Conference on Multiphase Flow (ICMF 2001, Paper No. 910). To understand turbulence in risers, measurements were done in the IIT riser with 530 micron glass beads using a PIV technique. This report summarizes the measurements and simulations completed so far. This work will continue under the sponsorship of the National Science Foundation and Dow Corning Corporation. This phase of the work is part of the DOE/Industry/University Multiphase Fluid Dynamics Research Consortium. Optimization of the LaPorte pilot plant reactor was attempted by rearranging the heat exchangers. The paper accepted for presentation at the Sixth World Congress of Chemical Engineering, Melbourne, Australia, September 23-27, 2001 is a part of this report.
Paul C.K. Lam; Isaac K. Gamwo; Dimitri Gidaspow
Experimental Parameters for a Reactor Antineutrino Experiment at Very Short Baselines
Reactor antineutrinos are used to study neutrino oscillation, search for signatures of non-standard neutrino interactions, and to monitor reactor operation for safeguard applications. The flux and energy spectrum of reactor antineutrinos can be predicted from the decays of the nuclear fission products. A comparison of recent reactor calculations with past measurements at baselines of 10-100m suggests a 5.7% deficit. Precision measurements of reactor antineutrinos at very short baselines O(1-10 m) can be used to probe this anomaly and search for possible oscillations into sterile neutrino species. This paper studies the experimental requirements for a new reactor antineutrino measurement at very short baselines and calculates the sensitivity of various scenarios. We conclude that an experiment at a typical research reactor provides 5{\\sigma} discovery potential for the favored oscillation parameter space with 3 years of data collection.
K. M. Heeger; B. R. Littlejohn; H. P. Mumm; M. N. Tobin
The Japan Times Printer Friendly Articles France has won the competition to host the International Thermonuclear Experimental Reactor (ITER), the world's first nuclear-
of securing a country's entire nuclear inventory, including decommissioned warheads and nuclear have led to major changes in the nuclear posture of the United States. The policy reviews undertaken alterations. As a result, the United States lacks a convincing rationale for its current nuclear force
Conceptual design description for the tritium recovery system for the US ITER (International Thermonuclear Experimental Reactor) Li sub 2 O/Be water cooled blanket
The tritium recovery system for the US ITER Li{sub 2}O/Be water cooled blanket processes two separate helium purge streams to recover tritium from the Li{sub 2}O zones and the Be zones of the blanket, to process the waste products, and to recirculate the helium back to the blanket. The components are selected to minimize the tritium inventory of the recovery system, and to minimize waste products. The system is robust to either an increase in the tritium release rate or to an in-leak of water in the purge system. Three major components were used to process these streams, first, 5A molecular sieves at {minus}196{degree}C separate hydrogen from the helium, second, a solid oxide electrolysis unit is used to reduce all molecular water, and third, a palladium/silver diffuser is used to ensure that only hydrogen (H{sub 2}, HT) species reach the cryogenic distillation unit. Other units are present to recover tritium from waste products but the three major components are the basis of the blanket tritium recovery system. 32 refs.
Finn, P.A.; Sze, D.K. (Argonne National Lab., IL (USA). Fusion Power Program); Clemmer, R.G. (Pacific Northwest Lab., Richland, WA (USA))
Asymmetric Explosions of Thermonuclear Supernovae
A type Ia supernova explosion starts in a white dwarf as a laminar deflagration at the center of the star and soon several hydrodynamic instabilities (in particular, the Rayleigh-Taylor (R-T) instability) begin to act. In previous work (Ghezzi, de Gouveia Dal Pino, & Horvath 2001), we addressed the propagation of an initially laminar thermonuclear flame in presence of a magnetic field assumed to be dipolar. We were able to show that, within the framework of a fractal model for the flame velocity, the front is affected by the field through the quenching of the R-T instability growth in the direction perpendicular to the field lines. As a consequence, an asymmetry develops between the magnetic polar and the equatorial axis that gives a prolate shape to the burning front. We have here computed numerically the total integrated asymmetry as the flame front propagates outward through the expanding shells of decreasing density of the magnetized white dwarf progenitor, for several chemical compositions, and found...
Ghezzi, C R; Horv�th, J E
A type Ia supernova explosion starts in a white dwarf as a laminar deflagration at the center of the star and soon several hydrodynamic instabilities (in particular, the Rayleigh-Taylor (R-T) instability) begin to act. In previous work (Ghezzi, de Gouveia Dal Pino, & Horvath 2001), we addressed the propagation of an initially laminar thermonuclear flame in presence of a magnetic field assumed to be dipolar. We were able to show that, within the framework of a fractal model for the flame velocity, the front is affected by the field through the quenching of the R-T instability growth in the direction perpendicular to the field lines. As a consequence, an asymmetry develops between the magnetic polar and the equatorial axis that gives a prolate shape to the burning front. We have here computed numerically the total integrated asymmetry as the flame front propagates outward through the expanding shells of decreasing density of the magnetized white dwarf progenitor, for several chemical compositions, and found that a total asymmetry of about 50 % is produced between the polar and equatorial directions for progenitors with a surface magnetic field B ? 5 � 10 7 G, and
C. R. Ghezzi; J. E. Horvath; S�o Paulo
A type Ia supernova explosion starts in a white dwarf as a laminar deflagration at the center of the star and soon several hydrodynamic instabilities (in particular, the Rayleigh-Taylor (R-T) instability) begin to act. In previous work (Ghezzi, de Gouveia Dal Pino, & Horvath 2001), we addressed the propagation of an initially laminar thermonuclear flame in presence of a magnetic field assumed to be dipolar. We were able to show that, within the framework of a fractal model for the flame velocity, the front is affected by the field through the quenching of the R-T instability growth in the direction perpendicular to the field lines. As a consequence, an asymmetry develops between the magnetic polar and the equatorial axis that gives a prolate shape to the burning front. We have here computed numerically the total integrated asymmetry as the flame front propagates outward through the expanding shells of decreasing density of the magnetized white dwarf progenitor, for several chemical compositions, and found that a total asymmetry of about 50 % is produced between the polar and equatorial directions for progenitors with a surface magnetic field B ~ 5 x 10^{7} G, and a composition C12 = 0.2 and O16 = 0.8 (in this case, the R-T instability saturates at scales \\~ 20 times the width of the flame front). This asymmetry is in good agreement with the inferred asymmetries from spectropolarimetric observations of very young supernova remnants, which have recently revealed intrinsic linear polarization interpreted as evidence of an asymmetric explosion in several objects,such as SN1999by, SN1996X, and SN1997dt. Larger magnetic field strengths will produce even larger asymmetries. We have also found that for lighter progenitors the total asymmetry is larger.
C. R. Ghezzi; E. M. de Gouveia Dal Pino; J. E. Horvath
A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling
Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.
Koch, M.; Kazimi, M.S.
An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory
Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.
Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))
Experimental characterization of slurry bubble-column reactor hydrodynamics
Sandia`s program to develop, implement, and apply diagnostics for hydrodynamic characterization of slurry bubble column reactors (SBCRs) at industrially relevant conditions is discussed. Gas liquid flow experiments are performed on an industrial scale. Gamma densitometry tomography (GDT) is applied to measure radial variations in gas holdup at one axial location. Differential pressure (DP) measurements are used to calculate volume averaged gas holdups along the axis of the vessel. The holdups obtained from DP show negligible axial variation for water but significant variations for oil, suggesting that the air water flow is fully developed (minimal flow variations in the axial direction) but that the air oil flow is still developing at the GDT measurement location. The GDT and DP gas holdup results are in good agreement for the air water flow but not for the air oil flow. Strong flow variations in the axial direction may be impacting the accuracy of one or both of these techniques. DP measurements are also acquired at high sampling frequencies (250 Hz) and are interpreted using statistical analyses to determine the physical mechanism producing each frequency component in the flow. This approach did not yield the information needed to determine the flow regime in these experiments. As a first step toward three phase material distribution measurements, electrical impedance tomography (EIT) and GDT are applied to a liquid solid flow to measure solids holdup. Good agreement is observed between both techniques and known values.
Shollenberger, K.A.; Torczynski, J.R.; Jackson, N.B.; O`Hern, T.J.
Experimental and Analytical Simulation of MFCI (Molten Fuel Coolant Interaction) during CDA (Core Disruptive Accident) in Sodium Cooled Fast Reactor.
??With increasing demand for understanding Severe Accident Scenario in Sodium Cooled Fast Reactors, there is an urgent need of enhancing numerical and experimental simulation techniques.� (more)
Natarajan, Venkataraman
First Thermonuclear Device Successfully Tested | National Nuclear Security
Thermonuclear Device Successfully Tested | National Nuclear Security Thermonuclear Device Successfully Tested | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > About Us > Our History > NNSA Timeline > First Thermonuclear Device Successfully Tested First Thermonuclear Device Successfully Tested December 31, 1952 Enewetak Atoll First Thermonuclear Device Successfully Tested
Experimental determination of neutron lifetimes through macroscopic neutron noise in the IPEN/MB-01 reactor
The neutron lifetimes of the core, reflector, and global were experimentally obtained through macroscopic neutron noise in the IPEN/MB-01 reactor for five levels of subcriticality. The theoretical Auto Power Spectral Densities were derived by point kinetic equations taking the reflector effect into account, and one of the approaches consider an additional group of delayed neutrons.
Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitaria, Sao Paulo (Brazil)
The objective if this study was to develop a predictive experimentally verified computational fluid dynamics (CFD) model for gas-liquid-solid flow. A three dimensional transient computer code for the coupled Navier-Stokes equations for each phase was developed. The principal input into the model is the viscosity of the particulate phase which was determined from a measurement of the random kinetic energy of the 800 micron glass beads and a Brookfield viscometer. The computed time averaged particle velocities and concentrations agree with PIV measurements of velocities and concentrations, obtained using a combination of gamma-ray and X-ray densitometers, in a slurry bubble column, operated in the bubbly-coalesced fluidization regime with continuous flow of water. Both the experiment and the simulation show a down-flow of particles in the center of the column and up-flow near the walls and nearly uniform particle concentration. Normal and shear Reynolds stresses were constructed from the computed instantaneous particle velocities. The PIV measurement and the simulation produced instantaneous particle velocities. The PIV measurement and the simulation produced similar nearly flat horizontal profiles of turbulent kinetic energy of particles. This phase of the work was presented at the Chemical Reaction Engineering VIII: Computational Fluid Dynamics, August 6-11, 2000 in Quebec City, Canada. To understand turbulence in risers, measurements were done in the IIT riser with 530 micron glass beads using a PIV technique. The results together with simulations will be presented at the annual meeting of AIChE in November 2000.
Paul Lam; Dimitri Gidaspow
EXPERIMENTAL BERYLLIUM OXIDE REACTOR PROGRAM. Quarterly Progress Report for the Period, January 1 through March 31, 1962
Progress made in the development of the Experimental Beryllium Oxide Reactor (EBOR) is reported. The objective of the EBOR program is to develop a gas-cooled, beryllium oxide-moderated reactor which can be used in conjunction with a closed-cycle gas turbine or a steam cycle for a small land-based or a maritime power plant. Progress is reported on reactor development, reactor physics, and materials development. (N.W.R.)
Current status of experimental breeder reactor-II [EBR-II] shutdown planning
The Experimental Breeder Reactor--II (EBR-II) at Argonne National Laboratory--West (ANL-W) in Idaho, was shutdown in September, 1994 as mandated by the US Department of Energy. This sodium cooled reactor had been in service since 1964, and was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the Sodium Process Facility. The sodium environment and the EBR-II configuration, combined with the radiation and contamination associated with thirty years of reactor operation, posed problems specific to liquid metal reactor deactivation. The methods being developed and implemented at EBR-II can be applied to other similar situations in the US and abroad.
McDermott, M. D.; Griffin, C. D.; Michelbacher, J. A.; Earle, O. K.
Computational and experimental prediction of dust production in pebble bed reactors -- Part I
This paper describes the computational modeling and simulation, and experimental testing of graphite moderators in frictional contacts as anticipated in a pebble bed reactor. The potential of carbonaceous particulate generation due to frictional contact at the surface of pebbles and the ensuing entrainment and transport into the gas coolant are safety concerns at elevated temperatures under accident scenarios such as air ingress in the high temperature gas-cooled reactor. The safety concerns are due to the documented ability of carbonaceous particulates to adsorb fission products and transport them in the primary circuit of the pebble bed reactor, thus potentially giving rise to a relevant source term under accident scenarios. Here, a finite element approach is implemented to develop a nonlinear wear model in air environment. In this model, material wear coefficient is related to the changes in asperity height during wear. The present work reports a comparison between the finite element simulations and the experimental results obtained using a custom-designed tribometer. The experimental and computational results are used to estimate the quantity of nuclear grade graphite dust produced from a typical anticipated configuration. In Part II, results from a helium environment at higher temperatures and pressures are experimentally studied.
Maziar Rostamian; Gannon Johnson; Mie Hiruta; Gabriel P. Potirniche; Abderrafi M. Ougouag; Joshua J. Cogliati; Akira Tokuhiro
Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor
In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs.
Boing, L.E.
Energy.gov (U.S. Department of Energy (DOE)) Indexed Site
Pre-Developmental Pre-Developmental INL EBR-II Wash Water Treatment Technologies (PBS # ADSHQTD0100 (0003199)) EBR-II Wash Water Workshop - The majority of the sodium has been removed, remaining material is mostly passivated. Similar closure projects have been successfully completed. Engineering needs to be developed to apply the OBA path. Page 1 of 2 Idaho National Laboratory Idaho Experimental Breeder Reactor-II Primary Tank System Wash Water Workshop Challenge In 1994 Congress ordered the shutdown of the Experimental Breeder Reactor-II (EBR-II) and a closure project was initiated. The facility was placed in cold shutdown, engineering began on sodium removal, the sodium was drained in 2001 and the residual sodium chemically passivated to render it less reactive in 2005. Since that time, approximately 700 kg of metallic sodium and 3500 kg of sodium bicarbonate remain in the facility. The
Theoretical and experimental studies of fixed-bed coal gasification reactors. Final report
A laboratory fixed-bed gasification reactor was designed and built with the objective of collecting operational data for model validation and parameter estimation. The reactor consists of a 4 inch stainless steel tube filled with coal or char. Air and steam is fed at one end of the reactor and the dynamic progress of gasification in the coal or char bed is observed through thermocouples mounted at various radial and axial locations. Product gas compositions are also monitored as a function of time. Results of gasification runs using Wyoming coal are included in this report. In parallel with the experimental study, a two-dimensional model of moving bed gasifiers was developed, coded into a computer program and tested. This model was used to study the laboratory gasifier by setting the coal feed rate equal to zero. The model is based on prior work on steady state and dynamic modeling done at Washington University and published elsewhere in the literature. Comparisons are made between model predictions and experimental results. These are also included in this report. 23 references, 18 figures, 6 tables.
Joseph, B.; Bhattacharya, A.; Salam, L.; Dudukovic, M.P.
>An assembly was constructed which simulated the moderator coolant annulus in the Experimental Gas-Cooled Reactor. This assembly was preheated to various temperatures and air was passed through the coolant annulus. Under certain conditions it was demonstrated that self-sustained combustion of the graphite could occur. Rates of temperature rise were generally less than 1 deg C/ min until the graphite temperature exceeded 700 deg C and then rise rates became very high. In these cxperimcnts, the assembly was not operated in such a manner as to give ignition temperatures eharaeteristic of the EGCR. These tests were designed only to investigate the effects of changing such parameters as the rate of coolant flow, air humidity, prior oxidation on the graphite, and air inlet temperature. A later series of experiments to more closely duplicate the EGCR operating conditions was completed and will be reported in a second report, HW- 71182. (auth)
Dahl, R.E.
Experimental power density distribution benchmark in the TRIGA Mark II reactor
In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)
Snoj, L.; Stancar, Z.; Radulovic, V.; Podvratnik, M.; Zerovnik, G.; Trkov, A. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Barbot, L.; Domergue, C.; Destouches, C. [CEA DEN, DER, Instrumentation Sensors and Dosimetry laboratory Cadarache, F-13108 Saint-Paul-Lez-Durance (France)
Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.
REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory
Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor
Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.
Pearson, J. Boise; Stewart, Eric T. [NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States); Reid, Robert S. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States)
Diffusion instability in a thermonuclear burn
Time-dependent, one-dimesional equations are solved to demonstrate the possibility of a superheating-diffusion instability in a steady-state fusion reactor.
Vasil' ev, N.N.; Lukash, V.E.; Nedospasov, A.V.
A simulated countercurrent moving-bed chromatographic reactor for the oxidative coupling of methane: Experimental results
The oxidative coupling reaction of methane (OCM) represents a potential commercial ethylene production route. However, the highest reported yields do not exceed 20%. The methane coupling reaction is accompanied by the undesired conversion of methane to carbon oxides. The relative amount of oxygen and methane along with other parameters, including temperature, determine the favored reaction pathway. High hydrocarbon to oxygen feed ratios give high ethane/ethylene selectivities but at the expense of the hydrocarbon conversion. When the methane to oxygen feed ratio is low, combustion is favored. The simulated countercurrent moving-bed chromatographic reactor (SCMCR) is applied to the OCM. A modified experimental configuration is designed and evaluated. A four-section apparatus, each containing a reaction and two separation columns, is used to quickly separate the reactants and products using the principles of simulated countercurrent flow. Simultaneous reaction and separation of the reactive products column is desired, but unattainable because of an incompatibility between OCM reaction and separation temperatures. Microreactor yields with a samarium oxide catalyst gives yields between 2% and 10%. Yields as high as 50% are observed with the same catalyst and run conditions in the SCMCR. These yields are significantly higher than previously reported values. The effects of temperature, feed switching time, and methane to oxygen feed ratio have been investigated. The reactor, while not fully optimized, does give promise as an alternative production method for ethylene.
Tonkovich, A.L.Y. [Pacific Northwest Lab., Richland, WA (United States); Carr, R.W. [Univ. of Minnesota, Minneapolis, MN (United States). Dept. of Chemical Engineering and Materials Sciences
Pure tension superconducting toroidal-field coil system design studies for the Argonne Experimental Power Reactor
As part of the Argonne Tokamak Experimental Power Reactor (TEPR) design studies, a toroidal field (TF) coil system has been designed. NbTi was chosen as the most suitable superconductor and 8T was regarded as a practical peak field level in this study. The 16-coil design was chosen as a reasonable compromise between 2 percent field ripple and 3 m access gap. To minimize the coil structure and the bending moments on the conductor, a pure tension coil shape is necessary. A correct approach for determining the pure tension coil profile in a bumpy TF coil system is given. Verification of the pure tension coil by a three- dimensional stress analysis is presented. For coil quench protection, a series- connected scheme is proposed. (auth)
Wang, S.T.; Purcell, J.R.; Demichele, D.W.; Turner, L.R.
The effects of seismic vibrations on the dynamic behavior of a composite system were analyzed. The equations of motion were derived and soIved with special emphasis on determining the resulting stresses. The method of analysis thus developed was applied to the composite structure consisting of the core, pressure vessel, and supporting skirt of the Experimental Gas-Cooled Reactor (EGCR). A system with three degrees of freedom was considered in order to determine the effects of an earthquake of the maximum intensity expected in the area surrounding Oak Ridge, Tennessee. The system of equations of motion was solved both numerically and analytically, and the resonant frequencies were determined. The seismic effect was shown to be small when the frequency of the seismic disturbance coincided with a natural frequency of the system. In particular, the shear stresses in the graphite core were shown to be negligible. (auth)
Witt, F.J.; Carver, D.R.; Maxwell, R.L.
Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power
This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power on spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.
Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.)
Experimental and Computational Study of the Flux Spectrum in Materials Irradiation Facilities of the High Flux Isotope Reactor
This report compares the available experimental neutron flux data in the High Flux Isotope Reactor (HFIR) to computational models of the HFIR loosely based on the experimental loading of cycle 400. Over the last several decades, many materials irradiation experiments have included fluence monitors which were subsequently used to reconstruct a coarse-group energy-dependent flux spectrum. Experimental values for thermal and fast neutron flux in the flux trap about the midplane are found to be 1.78 0.27 and 1.05 0:06 1E15 n/cm sec, respectively. The reactor physics code MCNP is used to calculate neutron flux in the HFIR at irradiation locations. The computational results are shown to correspond to closely to experimental data for thermal and fast neutron flux with calculated percent differences ranging from 0:55 13.20%.
McDuffee, Joel Lee [ORNL; Daly, Thomas F [ORNL
Study of Thermonuclear Alfven Instabilities in Next Step Burning Plasma Experiments
A study is presented for the stability of alpha-particle driven shear Alfven Eigenmodes (AE) for the normal parameters of the three major burning plasma proposals, ITER (International Thermonuclear Experimental Reactor), FIRE (Fusion Ignition Research Experiment), and IGNITOR (Ignited Torus). A study of the JET (Joint European Torus) plasma, where fusion alphas were generated in tritium experiments, is also included to attempt experimental validation of the numerical predictions. An analytic assessment of Toroidal AE (TAE) stability is first presented, where the alpha particle beta due to the fusion reaction rate and electron drag is simply and accurately estimated in 7-20 keV plasma temperature regime. In this assessment the hot particle drive is balanced against ion-Landau damping of the background deuterons and electron collision effects and stability boundaries are determined. Then two numerical studies of AE instability are presented. In one the High-n stability code HINST is used . This code is capable of predicting instabilities of low and moderately high frequency Alfven modes. HINST computes the non-perturbative solution of the Alfven eigenmodes including effects of ion finite Larmor radius, orbit width, trapped electrons etc. The stability calculations are repeated using the global code NOVAK. We show that for these tokamaks the spectrum of the least stable AE modes are TAE that appear at medium-/high-n numbers. In HINST TAEs are locally unstable due to the alphas pressure gradient in all the devices under the consideration except IGNITOR. However, NOVAK calculations show that the global mode structure enhances the damping mechanisms and produces stability in all configurations considered here. A serious question remains whether the perturbation theory used in NOVAK overestimates the stability predictions, so that it is premature to conclude that the nominal operation of all three proposals are stable to AEs. In addition NBI ions produce a strong stabilizing effect for JET. However, in ITER the beam energies needed to penetrate to the core must be high so that a diamagnetic drift frequency comparable to that of the alpha particles is produced by the beam ions which induces a destabilizing effect.
N.N. Gorelenkov; H.L. Berk; R. Budny; C.Z. Cheng; G.-Y. Fu; W.W. Heidbrink; G. Kramer; D. Meade; and R. Nazikian
Experimental and Numerical Investigation of Boron Dilution Transients in Pressurized Water Reactors
Within the pressurized water reactor (PWR) safety analyses, attention has increasingly focused in recent years on boron dilution events that could potentially lead to reactivity transients. Mixing of the low-boron water with the ambient coolant of higher boron content provides an important mitigation mechanism before the low-boron water enters the core.Experimental support is needed to validate the computational tools to be applied to analyze the mixing of the low-boron water. Experiments were performed in the three test facilities - the Upper Plenum Test Facility (UPTF), the Primaerkreislauf (PKL), and the Rossendorf coolant mixing model (ROCOM) - in Germany.The relevant PKL and UPTF tests were focused on small-break loss-of-coolant accident (SBLOCA) scenarios with reflux-condenser mode and restart of natural circulation. The two test facilities represent a typical western-type PWR and are/were operated by Siemens/KWU now Framatome ANP in Germany. While the restart of natural circulation was investigated in the PKL system test facility (volume 1:145, height 1:1), the UPTF experiments dealt with the mixing of water flows with different boron concentration in the cold legs, reactor pressure vessel (RPV) downcomer, and the lower plenum (all these components were full-scale models).The results from the PKL test facility demonstrate that in case of a postulated SBLOCA with reflux condensation phase, natural circulation does not start up simultaneously in all loops. This means that slugs of condensate, which might have accumulated in the pump seal during reflux-condenser mode of operation, would reach the RPV at different points in time. The UPTF tests showed an almost ideal mixing of water flows with different boron concentration in the RPV downcomer.The ROCOM test facility has been built in a linear scale of 1:5 for the investigation of coolant mixing phenomena in a wide range of flow conditions in the RPV of the German KONVOI-type PWR. The test results presented are focused on the mixing of a slug of deborated water during the startup of the first reactor coolant pump. Based on experimentally determined pulse responses, a semianalytical model for the description of coolant mixing inside the KONVOI RPV has been developed. Calculations for a presumed boron dilution event during the startup of the first reactor coolant pump have been carried out by means of the semianalytical model and independently by means of the computational fluid dynamics code CFX-4. The semianalytical model is able to describe the time dependent behavior of the deboration front at each fuel element position in a good agreement with the experiment. All main mixing effects, observed in the experiment, are also reproduced by the CFX calculation.
Hertlein, Roland J. [Framatome ANP GmbH (France); Umminger, Klaus [Framatome ANP GmbH (France); Kliem, Soeren [Forschungszentrum Rossendorf e.V. (Germany); Prasser, Horst-Michael [Forschungszentrum Rossendorf e.V. (Germany); Hoehne, Thomas [Forschungszentrum Rossendorf e.V. (Germany); Weiss, Frank-Peter [Forschungszentrum Rossendorf e.V. (Germany)
Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor
Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.
Richard Schultz
Experimental techniques for hydrodynamic characterization of multiphase flows in slurry-phase bubble-column reactors
Slurry-phase bubble-column Fischer-Tropsch (FT) reactors are recognized as one of the more promising technologies for converting synthesis gas from coal into liquid fuel products (indirect liquefaction). However, hydrodynamic effects must be considered when attempting to scale these reactors to sizes of industrial interest. The objective of this program is to facilitate characterization of reactor hydrodynamics by developing and applying noninvasive tomographic diagnostics capable of measuring gas holdup spatial distribution in these reactors.
Torczynski, J.R.; O`Hern, T.J.; Adkins, D.R.; Shollenberger, K.A.; Mondy, L.A.; Jackson, N.B.
Experimental study of the oxidation of methyl oleate in a jet-stirred reactor
The experimental study of the oxidation of a blend containing n-decane and a large unsaturated ester, methyl oleate, was performed in a jet-stirred reactor over a wide range of temperature covering both low and high temperature regions (550-1100 K), at a residence time of 1.5 s, at quasi atmospheric pressure with high dilution in helium (n-decane and methyl oleate inlet mole fractions of 1.48 x 10{sup -3} and 5.2 x 10{sup -4}) and under stoichiometric conditions. The formation of numerous reaction products was observed. At low and intermediate temperatures, the oxidation of the blend led to the formation of species containing oxygen atoms like cyclic ethers, aldehydes and ketones deriving from n-decane and methyl oleate. At higher temperature, these species were not formed anymore and the presence of unsaturated species was observed. Because of the presence of the double bond in the middle of the alkyl chain of methyl oleate, the formation of some specific products was observed. These species are dienes and esters with two double bonds produced from the decomposition paths of methyl oleate and some species obtained from the addition of H-atoms, OH and HO{sub 2} radicals to the double bond. Experimental results were compared with former results of the oxidation of a blend of n-decane and methyl palmitate performed under similar conditions. This comparison allowed highlighting the similarities and the differences in the reactivity and in the distribution of the reaction products for the oxidation of large saturated and unsaturated esters. (author)
Bax, Sarah; Hakka, Mohammed Hichem; Glaude, Pierre-Alexandre; Herbinet, Olivier; Battin-Leclerc, Frederique [Laboratoire Reactions et Genie des Procedes, Nancy Universite, CNRS, ENSIC, BP 20451, 1 rue Grandville, 54001 Nancy (France)
Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor
This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.
Ilas, Germina [ORNL; Gauld, Ian C [ORNL
Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory
The Final Report for the Decontamination and Decommissioning (D&D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D&D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D&D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D&D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a {open_quotes}Radiologically Controlled Area,{close_quotes} noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion).
Fellhauer, C.R.; Boing, L.E. [Argonne National Lab., IL (United States); Aldana, J. [NES, Inc., Danbury, CT (United States)
A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.
Roman, W.G.
Experimental and simulated dosimetry of the University of Utah triga reactor.
??Simulated neutron and gamma transport enable the gamma dose to be estimated at the surface of the University of Utah TRIGA Reactor UUTR pool. These� (more)
Marble, Benjamin James
Prospects for Tokamak Fusion Reactors
This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.
Sheffield, J.; Galambos, J.
Short Term Irradiation Test of Fuel Containing Minor Actinides Using the Experimental Fast Reactor Joyo
A mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast rector Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted as part of the short-term phase of this program in May and August 2006. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), and MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX). The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes. After 10 minutes irradiation test, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins with neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. The linear heat rate for each MA-MOX test fuel pin was calculated using the Monte Carlo calculation code MCNP. The calculated fission rates were compared with the measured data based on the Nd-148 method. The maximum linear heat rate was approximately 444{+-}19 W/cm at the actual reactor power of 119.6 MWt. Post irradiation examination of these pins to confirm the absence of fuel melting and the local concentration under irradiation of NpO{sub 2-x} or AmO{sub 2-x}, in the (U,Pu)0{sub 2-x}, fuel are underway. The test results are expected to reduce uncertainties on the margin in the thermal design for MA-MOX fuel. (authors)
Sekine, Takashi; Soga, Tomonori; Koyama, Shin-ichi; Aoyama, Takafumi [Oarai Research and Development Center, Japan Atomic Energy Agency. 4002 Narita, Oarai, Ibaraki 311-1393 (Japan); Wootan, David [Pacific Northwest National Laboratoy, M/S K8-34, P.O. Box 999 Richland, WA 99352 (United States)
Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor
The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.
S.T. Revankar; W. Zhou; Gavin Henderson
The KEWB Program is devoted to the study of the dynamic behavior of homogeneous type research reactors. The objectives of this program include studies to develop better and more complete understanding of phenomena which contribute to the kinetic behavior and the inherent safety of the water boiler reactor. The approach to the objectives has heen to construct a prototype 50 kw homogeneous reactor with the necessary auxiliary apparatus and to study the transient behavior of the system as a function of the more significant parameters which affect this behavior. These include the amount of reactivity release, rate of reactivity release, initial core pressure, initial core temperature, initial reactor power, and void volume above the core. Data are plotted. (auth)
Remley, M.E.; Flora, J.W.; Hetrick, D.L.; Muller, D.R.; Gardner, E.L.; Wimmer, R.E.; Stitt, R.K.; Gamble, D.P.
Power transient analyses of experimental in-reflector devices during safety shutdown in Jules Horowitz Reactor (JHR)
The Jules Horowitz Reactor (JHR) is designed to be a 100 MW material testing reactor (MTR) and it is expected to become the reference facility in the framework of European nuclear research activity. As the core neutron spectrum is quite fast, several experimental devices concerning fuel studies have been conceived to be placed in the reflector in order to exploit a proper thermal neutron flux irradiation. Since the core power is relatively high, the neutronic coupling between the reactor core and the reflector devices has to be taken into account for different rod insertions. In fact the thermal power produced within the fuel samples is considerable. Heat removal during shutdown is a main topic in nuclear safety and it is worth to analyse thermal power transients in fuel samples as well. Here a thermal hydraulic model for JHR core is proposed aiming at a simple and representative description as far as reactivity feedbacks are concerned. Then it is coupled with a neutronic pointwise kinetics analysis by means of the DULCINEE code to compute core power transient calculations. Moreover, some reflector-core coupling evaluations are performed through Monte Carlo method using the TRIPOLI 4.7 code. The JHR equilibrium cycle is considered with respect to four fuel compositions namely Beginning of Cycle (BOC), Xenon Saturation Point (XSP), Middle of Cycle (MOC) and End of Cycle (EOC). Then thermal power transients in the experimental reflector devices are evaluated during safety shutdowns and they are verified for all these cycle steps. (authors)
Camprini, P. C.; Sumini, M. [Univ. of Bologna (Italy); Artioli, C. [National Agency for New Technologies, Energy and Sustainable Economic Development ENEA (Italy); Gonnier, C.; Pouchin, B.; Bourdon, S. [Atomic Energy Commission CEA (France)
Considerable progress has been achieved in understanding three-phase reactors from the point of view of kinetic theory. In a paper in press for publication in Chemical Engineering Science (Wu and Gidaspow, 1999) we have obtained a complete numerical solution of bubble column reactors. In view of the complexity of the simulation a better understanding of the processes using simplified analytical solutions is required. Such analytical solutions are presented in the attached paper, Large Scale Oscillations or Gravity Waves in Risers and Bubbling Beds. This paper presents analytical solutions for bubbling frequencies and standing wave flow patterns. The flow patterns in operating slurry bubble column reactors are not optimum. They involve upflow in the center and downflow at the walls. It may be possible to control flow patterns by proper redistribution of heat exchangers in slurry bubble column reactors. We also believe that the catalyst size in operating slurry bubble column reactors is not optimum. To obtain an optimum size we are following up on the observation of George Cody of Exxon who reported a maximum granular temperature (random particle kinetic energy) for a particle size of 90 microns. The attached paper, Turbulence of Particles in a CFB and Slurry Bubble Columns Using Kinetic Theory, supports George Cody's observations. However, our explanation for the existence of the maximum in granular temperature differs from that proposed by George Cody. Further computer simulations and experiments involving measurements of granular temperature are needed to obtain a sound theoretical explanation for the possible existence of an optimum catalyst size.
Isaac K. Gamwo; Dimitri Gidaspow
Method of achieving the controlled release of thermonuclear energy
A method of achieving the controlled release of thermonuclear energy by illuminating a minute, solid density, hollow shell of a mixture of material such as deuterium and tritium with a high intensity, uniformly converging laser wave to effect an extremely rapid build-up of energy in inwardly traveling shock waves to implode the shell creating thermonuclear conditions causing a reaction of deuterons and tritons and a resultant high energy thermonuclear burn. Utilizing the resulting energy as a thermal source and to breed tritium or plutonium. The invention also contemplates a laser source wherein the flux level is increased with time to reduce the initial shock heating of fuel and provide maximum compression after implosion; and, in addition, computations and an equation are provided to enable the selection of a design having a high degree of stability and a dependable fusion performance by establishing a proper relationship between the laser energy input and the size and character of the selected material for the fusion capsule.
Brueckner, Keith A. (Ann Arbor, MI)
Automated operator procedure prompting for startup of Experimental Breeder Reactor-2
This report describes the development of an operator procedure prompting aid for startup of a nuclear reactor. This operator aid is a preliminary design for a similar aid that eventually will be used with the Advanced Liquid Metal Reactor (ALMR) presently in the design stage. Two approaches were used to develop this operator procedure prompting aid. One method uses an expert system software shell, and the other method uses database software. The preliminary requirements strongly pointed toward features traditionally associated with both database and expert systems software. Database software usually provides data manipulation flexibility and user interface tools, and expert systems tools offer sophisticated data representation and reasoning capabilities. Both methods, including software and associated hardware, are described in this report. Proposals for future enhancements to improve the expert system approach to procedure prompting and for developing other operator aids are also offered. 25 refs., 14 figs.
Renshaw, A.W.; Ball, S.J.; Ford, C.E.
Experimental and design experience with passive safety features of liquid metal reactors
Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-II with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions.
Lucoff, D.M.; Waltar, A.E. [Westinghouse Hanford Co., Richland, WA (United States); Sackett, J.I. [Argonne National Lab., Idaho Falls, ID (United States); Salvatores, M. [CEA, 75 - Paris (France); Aizawa, K. [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)
Experimental investigation into fast pyrolysis of biomass using an entrained-flow reactor
Pyrolysis experiments were performed using 30 and 90cm entrained-flow reactors, with steam as a carrier gas and two different feedstocks - wheat straw and powdered material drived from municipal solid waste (ECO-II TM). Reactor wall temperature was varied from 700/sup 0/ to 1400/sup 0/C. Gas composition data from the ECO-II tests were comparable to previously reported data but ethylene yield appeared to vary with reactor wall temperature and residence time. The important conclusion from the wheat straw tests is that olefin yields are about one half that obtained from ECO-II. Evidence was found that high olefin yields from ECO-II are due to the presence of plastics in the feedstock. Batch experiments were run on wheat straw using a Pyroprobe/sup TM/. The samples were heated at a high rate (20,000/sup 0/ C/sec) to 1000/sup 0/ and held at 1000/sup 0/C for a variable period of time from 0.05 to 4.95s. For times up to 0.15s volume fractions of ethylene, propylene, and methane increase while that of carbon dioxide decreases. Subsequently, only carbon monoxide and hydrogen are produced. The change may be related to poor thermal contact and suggests caution in using the Pyroprobe.
Bohn, M.; Benham, C.
Experimental characterization of an Ion Transport Membrane (ITM) reactor for methane oxyfuel combustion
Ion Transport Membranes (ITM) which conduct both electrons and oxygen ions have been investigated experimentally for oxygen separation and fuel (mostly methane) conversion purposes over the last three decades. The fuel ...
Apo, Daniel Jolomi
Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics
This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.
Tokuhiro, Akira; Jones, Byron
Design considerations in inertially-confined fusion reactors
This paper discusses the effects of short time pulses of energetic particles and waves typical of inertially-confined thermonuclear reactions on the first wall, blanket and shield of conceptual reactors. Several reactor designs are presented which attempt to cope with the various problems from the microexplosion debris. Fusion-fission hybrid reactors are also discussed. Emphasis is placed on the first-wall problems of laser-initiated, inertially confined fusion reactors using the deuterium-tritium fuel cycle.
Hovingh, J.
Fusion reactors: a remote possibility
The next generation of controlled thermonuclear reactor experiments will be faced with the handling problems of tritium and neutron activation that will dominate the safety and maintenance problems of future fusion reactors. The nuclear industry has been working with highly radioactive systems for many years and has developed the tools and methods to do safely productive work in the presence of high radiation fields. These methods can be applied to CTR work by extending them to the unique problems associated with fusion reactors. (auth)
Doggett, J.N.
Heat Transfer Simulation of Reactor Cavity Cooling System Experimental Facility using RELAP5-3D and Generation of View Factors using MCNP
As one of the most attractive reactor types, The High Temperature Gas-cooled Reactor (HTGR) is designed to be passively safe with the incorporation of Reactor Cavity Cooling System (RCCS). In this paper, a RELAP5-3D simulation model is set up based on the 1/16 scale experimental facility established by Texas A&M University. Also, RELAP5-3D input decks are modified to replicate the experiment procedures and the experimental results are compared with the simulation results. The results show there is a perfect match between experimental and simulation results. Radiation heat transfer dominates in the heat transfer process of high temperature gas-cooled reactor due to its high operation temperature. According to experimental research done with the RCCS facility in Texas A&M University, radiation heat transfer takes up 80% of the total heat transferred to standing pipes. In radiation heat transfer, the important parameters are view factors between surfaces. However, because of the geometrical complexity in the experimental facility, it is hard to use the numerical method or analytical view factor formula to calculate view factors. In this project, MCNP based on the Monte Carlo method is used to generate view factors for RELAP5-3D input. MCNP is powerful in setting up complicated geometry, source definition and tally application. In the end, RCCS geometry is set up using MCNP and view factors are calculated.
Wu, Huali
Interpretation of experimental data on safety parameters from standpoint of mathematical nuclear reactor theory
The paper deals with the problem of correct processing of experimental information for unambiguous determination of current reactivity value. The methodology for reactivity determination that is based on spectral decomposition of the solution obtained for non-stationary equation of neutron transport is presented. The proposed spectral projection algorithm takes into consideration the effects related with space-time re-distribution of neutron field and thus excludes the main systematic error and increases accuracy of current reactivity values. The paper presents the results of model calculations, demonstrating the effectiveness of the proposed method. (authors)
Cherezov, A. L.; Shchukin, N. V. [National Nuclear Research Univ. MEPhI, 115409, Kashirskoe, 31 (Russian Federation)
Occupational health physics at a fusion reactor
Future generation of electrical power using controlled thermonuclear reactors will involve both traditional and new concerns for health protection. A review of the problems associated with exposures to tritium and magnetic fields is presented with emphasis on the occupational worker. The radiological aspects of tritium, inventories and loss rates of tritium for fusion reactors, and protection of the occupational worker are discussed. Magnetic fields in which workers may be exposed routinely and possible biological effects are also discussed. (auth)
Shank, K.E.; Easterly, C.E.; Shoup, R.L.
Detonating Failed Deflagration Model of Thermonuclear Supernovae I. Explosion Dynamics
We present a detonating failed deflagration model of Type Ia supernovae. In this model, the thermonuclear explosion of a massive white dwarf follows an off-center deflagration. We conduct a survey of asymmetric ignition configurations initiated at various distances from the stellar center. In all cases studied, we find that only a small amount of stellar fuel is consumed during deflagration phase, no explosion is obtained, and the released energy is mostly wasted on expanding the progenitor. Products of the failed deflagration quickly reach the stellar surface, polluting and strongly disturbing it. These disturbances eventually evolve into small and isolated shock-dominated regions which are rich in fuel. We consider these regions as seeds capable of forming self-sustained detonations that, ultimately, result in the thermonuclear supernova explosion. Preliminary nucleosynthesis results indicate the model supernova ejecta are typically composed of about 0.1-0.25 Msun of silicon group elements, 0.9-1.2 Msun of iron group elements, and are essentially carbon-free. The ejecta have a composite morphology, are chemically stratified, and display a modest amount of intrinsic asymmetry. The innermost layers are slightly egg-shaped with the axis ratio ~1.2-1.3 and dominated by the products of silicon burning. This central region is surrounded by a shell of silicon-group elements. The outermost layers of ejecta are highly inhomogeneous and contain products of incomplete oxygen burning with only small admixture of unburned stellar material. The explosion energies are ~1.3-1.5 10^51 erg.
Tomasz Plewa
Power Supply Reliability Estimates for Experimental Fusion Facilities
This paper presents the results of a task to analyze the operating experience data for large, pulsed power supplies used at the DIII-D tokamak. This activity supports the International Thermonuclear Experimental Reactor (ITER) project by giving fusion-specific reliability values for large power supplies that energize neutral beams and magnets. These failure rate data are necessary to perform system availability calculations and to make estimates of the frequency of safety-significant events (e.g., power supply arcs or fires) that might occur in other fusion facilities such as ITER. The analysis shows that the DIII-D data results compare well with the results of similar data analysis work that the Italian National Agency for New Technologies, Energy and the Environment (ENEA) has performed on the JET tokamak and compare fairly with data from two accelerators.
Cadwallader, lee; Pinna, Tonio; Petersen, Peter
Cryogenic thermonuclear fuel implosions on the National Ignition Facility
The first inertial confinement fusion implosion experiments with equimolar deuterium-tritium thermonuclear fuel have been performed on the National Ignition Facility. These experiments use 0.17 mg of fuel with the potential for ignition and significant fusion yield conditions. The thermonuclear fuel has been fielded as a cryogenic layer on the inside of a spherical plastic capsule that is mounted in the center of a cylindrical gold hohlraum. Heating the hohlraum with 192 laser beams for a total laser energy of 1.6 MJ produces a soft x-ray field with 300 eV temperature. The ablation pressure produced by the radiation field compresses the initially 2.2-mm diameter capsule by a factor of 30 to a spherical dense fuel shell that surrounds a central hot-spot plasma of 50 {mu}m diameter. While an extensive set of x-ray and neutron diagnostics has been applied to characterize hot spot formation from the x-ray emission and 14.1 MeV deuterium-tritium primary fusion neutrons, thermonuclear fuel assembly is studied by measuring the down-scattered neutrons with energies in the range of 10 to 12 MeV. X-ray and neutron imaging of the compressed core and fuel indicate a fuel thickness of (14 {+-} 3) {mu}m, which combined with magnetic recoil spectrometer measurements of the fuel areal density of (1 {+-} 0.09) g cm{sup -2} result in fuel densities approaching 600 g cm{sup -3}. The fuel surrounds a hot-spot plasma with average ion temperatures of (3.5 {+-} 0.1) keV that is measured with neutron time of flight spectra. The hot-spot plasma produces a total fusion neutron yield of 10{sup 15} that is measured with the magnetic recoil spectrometer and nuclear activation diagnostics that indicate a 14.1 MeV yield of (7.5{+-}0.1) Multiplication-Sign 10{sup 14} which is 70% to 75% of the total fusion yield due to the high areal density. Gamma ray measurements provide the duration of nuclear activity of (170 {+-} 30) ps. These indirect-drive implosions result in the highest areal densities and neutron yields achieved on laser facilities to date. This achievement is the result of the first hohlraum and capsule tuning experiments where the stagnation pressures have been systematically increased by more than a factor of 10 by fielding low-entropy implosions through the control of radiation symmetry, small hot electron production, and proper shock timing. The stagnation pressure is above 100 Gbars resulting in high Lawson-type confinement parameters of P{tau} Asymptotically-Equal-To 10 atm s. Comparisons with radiation-hydrodynamic simulations indicate that the pressure is within a factor of three required for reaching ignition and high yield. This will be the focus of future higher-velocity implosions that will employ additional optimizations of hohlraum, capsule and laser pulse shape conditions.
Glenzer, S. H.; Callahan, D. A.; MacKinnon, A. J.; Alger, E. T.; Berger, R. L.; Bernstein, L. A.; Bleuel, D. L.; Bradley, D. K.; Burkhart, S. C.; Burr, R.; Caggiano, J. A.; Castro, C.; Choate, C.; Clark, D. S.; Celliers, P.; Cerjan, C. J.; Collins, G. W.; Dewald, E. L.; DiNicola, P.; DiNicola, J. M. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others
An Experimental Shield Test Facility for the Development of Minimum Weight Shields for Compact Reactor Power Systems
Discussions are given of the characteristics of fission-source plate, graphite reactor, and pool-type reactor facilities applicable to development studies of minimum weight shielding materials. Advantages of a proposed SNAP dual-purpose shielding facility are described in terms of a disk-shaped fission-source plate, reactor, and building. A program for the study of advanced shielding materials is discussed for materials and configuations to be evaluted with the fission-source plate, the testing of the prototype at high-power levels, and full-power tests on the actual reactor.
Tomlinson, R.L.
Compilation and evaluation of atomic and molecular data relevant to controlled thermonuclear research needs: USA programs
The U.S. role in the compilation and evaluation of atomic data for controlled thermonuclear research is discussed in the following three areas: (1) atomic structure data, (2) atomic collision data, and (3) surface data. (MOW)
Barnett, C.F.
applications of sensors and modeling to materials processing: iii - TMS
... based sensor is being developed for remote metrology of plasma facing material surfaces in the International Thermonuclear Experimental Reactor (ITER ).
Comparison and validation of HEU and LEU modeling results to HEU experimental benchmark data for the Massachusetts Institute of Technology MITR reactor.
The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Towards this goal, comparisons of MCNP5 Monte Carlo neutronic modeling results for HEU and LEU cores have been performed. Validation of the model has been based upon comparison to HEU experimental benchmark data for the MITR-II. The objective of this work was to demonstrate a model which could represent the experimental HEU data, and therefore could provide a basis to demonstrate LEU core performance. This report presents an overview of MITR-II model geometry and material definitions which have been verified, and updated as required during the course of validation to represent the specifications of the MITR-II reactor. Results of calculations are presented for comparisons to historical HEU start-up data from 1975-1976, and to other experimental benchmark data available for the MITR-II Reactor through 2009. This report also presents results of steady state neutronic analysis of an all-fresh LEU fueled core. Where possible, HEU and LEU calculations were performed for conditions equivalent to HEU experiments, which serves as a starting point for safety analyses for conversion of MITR-II from the use of HEU fuel to the use of UMo LEU fuel.
Newton, T. H.; Wilson, E. H; Bergeron, A.; Horelik, N.; Stevens, J. (Nuclear Engineering Division); (MIT Nuclear Reactor Lab.)
Thermonuclear Fusion Energy : Assessment and Next Step Ren Pellat
, will have to be properly considered in tokamak reactor design. MHD theory and modelling have been Atomic Energy Chairman of the CCE-FU Energy and environment. Nuclear and renewable energies 8 9 March allowed to continuously progress towards the fusion reactor which stays a physics and technology ambitious
Nucleosynthesis in type Ia supernovae driven by asymmetric thermonuclear ignition
Type Ia Supernovae (SNe Ia) are believed to be thermonuclear explosions of a white dwarf. They can be used as mature cosmological standardized candles, leading to the discovery of the accelerating expansion of the Universe. However, the explosion mechanism has not yet been fully clarified. In this paper, we first present nucleosynthetic features of a leading explosion scenario, namely a delayed-detonation scenario. Based on this, we propose a new and strong observational constraint on the explosion mechanism through emission lines from neutron-rich Fe-peaks. Especially, we show that an asymmetry in the explosion is likely a generic feature. We further argue that the diversity arising from various viewing angles can be an origin of observational diversities of SNe Ia seen in their spectral features (suspected possible biases in cosmology) and colors (related to the extinction estimate in cosmology). Using these new insights could open up a possibility of using SNe Ia as more precise distance indicators than currently employed.
Maeda, Keiichi [Institute for the Physics and Mathematics of the Universe (IPMU), Todai Institutes for Advanced Study (TODIAS), University of Tokyo, 5-1-5 Kashiwanoha, Kashiwa, Chiba 277-8583 (Japan)
Experimental Determination of the Effect of Reactor Radiation on the Thermal Conductivity of Uranium-Impregnated Graphite
Experiments are described in which the change in thermal conductivity of U-impregnated graphite under neutron irradiation was measured. Thermal resistivities relative to the thermal resistivity of undamaged impregnated graphite are reorted as functions of exposure. From applications of the expermental results to the North American Aviation low-power research reactor the peak tem. of the core is determined for a given reactor power and time of operation.
Hetrick, D.L.; McCarty, W.K.; Steele, G.N.; Brown, M.S.; Clark, E.V.; Holmes, F.R.; Howard, D.F.; McElroy, W.N.; Shields, B.L.
Pacific Northwest Laboratory report on controlled thermonuclear reactor technology, January 1975 - September 1975
The PNL staff has been studying fusion technology in areas such as economics, fusion-fission hybrid concepts, materials, neutronics, environment and safety. These studies have been scoped to make efficient use of ERDA resources, and to complement and support efforts at other laboratories. The effect the plasma and associated radiation and emission will have upon the surfaces of the first wall are being studied. Neutron sputtering experiments were made on niobium and gold and the results were evaluated for absolute neutron yields. Molybdenum and vanadium were studied for effects of ion bombardment under various conditions of helium injection. Graphite cloth is being irradiated for examination of radiation effects because it is suggested for use in several CTR concepts as a shield between the plasma and the first wall. Helium effects are being studied to characterize degradation of structural metal properties. Work is progressing on absolute measurement of the electrical resistivity of insulators and the demonstration of the feasibility of producing insulating coatings by sputter deposition. (auth)
Millihertz quasi-periodic oscillations and thermonuclear bursts from Terzan 5: A showcase of burning regimes
We present a comprehensive study of the thermonuclear bursts and millihertz quasi-periodic oscillations (mHz QPOs) from the neutron star (NS) transient and 11 Hz X-ray pulsar IGR J17480�2446, located in the globular cluster ...
Linares, M.
Composite Materials under Extreme Radiation and Temperature Environments of the Next Generation Nuclear Reactors
In the nuclear energy renaissance, driven by fission reactor concepts utilizing very high temperatures and fast neutron spectra, materials with enhanced performance that exceeds are expected to play a central role. With the operating temperatures of the Generation III reactors bringing the classical reactor materials close to their performance limits there is an urgent need to develop and qualify new alloys and composites. Efforts have been focused on the intricate relations and the high demands placed on materials at the anticipated extreme states within the next generation fusion and fission reactors which combine high radiation fluxes, elevated temperatures and aggressive environments. While nuclear reactors have been in operation for several decades, the structural materials associated with the next generation options need to endure much higher temperatures (1200 C), higher neutron doses (tens of displacements per atom, dpa), and extremely corrosive environments, which are beyond the experience on materials accumulated to-date. The most important consideration is the performance and reliability of structural materials for both in-core and out-of-core functions. While there exists a great body of nuclear materials research and operating experience/performance from fission reactors where epithermal and thermal neutrons interact with materials and alter their physio-mechanical properties, a process that is well understood by now, there are no operating or even experimental facilities that will facilitate the extreme conditions of flux and temperature anticipated and thus provide insights into the behaviour of these well understood materials. Materials, however, still need to be developed and their interaction and damage potential or lifetime to be quantified for the next generation nuclear energy. Based on material development advances, composites, and in particular ceramic composites, seem to inherently possess properties suitable for key functions within the operating envelope of both fission and fusion reactors. In advanced fission reactors composite materials are being designed in an effort to extend the life and improve the reliability of fuel rod cladding as well as structural materials. Composites are being considered for use as core internals in the next generation of gas-cooled reactors. Further, next-generation plasma-fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER) will rely on the capabilities of advanced composites to safely withstand extremely high neutron fluxes while providing superior thermal shock resistance.
Simos, N.
Design of an experimental loop for post-LOCA heat transfer regimes in a Gas-cooled Fast Reactor
The goal of this thesis is to design an experimental thermal-hydraulic loop capable of generating accurate, reliable data in various convection heat transfer regimes for use in the formulation of a comprehensive convection ...
Cochran, Peter A. (Peter Andrew)
Laser-fusion targets for reactors
A laser target comprising a thermonuclear fuel capsule composed of a centrally located quantity of fuel surrounded by at least one or more layers or shells of material for forming an atmosphere around the capsule by a low energy laser prepulse. The fuel may be formed as a solid core or hollow shell, and, under certain applications, a pusher-layer or shell is located intermediate the fuel and the atmosphere forming material. The fuel is ignited by symmetrical implosion via energy produced by a laser, or other energy sources such as an electron beam machine or ion beam machine, whereby thermonuclear burn of the fuel capsule creates energy for applications such as generation of electricity via a laser fusion reactor.
Nuckolls, John H. (Livermore, CA); Thiessen, Albert R. (Livermore, CA)
Effects of a liquid lithium curtain as the first wall in the Fusion Experimental Breeder (FEB-E) reactor plasma
Effects of a liquid lithium curtain as the first wall in the Fusion Experimental Breeder (FEB, IL 60439 USA Abstract We study the effects of a liquid lithium curtain used as the first wall the surface temperature of a liquid lithium curtain and the effective plasma charge, fuel dilution, and fusion
Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors
The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.
Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C. [and others
Three-dimensional neutronics optimization of helium-cooled blanket for multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)
Three-dimensional neutronics optimization calculations were performed to analyse the parameters of Tritium Breeding Ratio (TBR) and maximum average Power Density (PDmax) in a helium-cooled multi-functional experimental fusion-fission hybrid reactor named FDS (Fusion-Driven hybrid System)-MFX (Multi-Functional experimental) blanket. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this contribution, the most significant and main goal of the FDS-MFX blanket is to achieve the PDmax of about 100 MW/m3 with self-sustaining tritium (TBR {>=} 1.05) based on the second-stage test with uranium-fueled blanket to check and validate the demonstrator reactor blanket relevant technologies based on the viable fusion and fission technologies. Four different enriched uranium materials were taken into account to evaluate PDmax in subcritical blanket: (i) natural uranium, (ii) 3.2% enriched uranium, (iii) 19.75% enriched uranium, and (iv) 64.4% enriched uranium carbide. These calculations and analyses were performed using a home-developed code VisualBUS and Hybrid Evaluated Nuclear Data Library (HENDL). The results showed that the performance of the blanket loaded with 64.4% enriched uranium was the most attractive and it could be promising to effectively obtain tritium self-sufficiency (TBR-1.05) and a high maximum average power density ({approx}100 MW/m{sup 3}) when the blanket was loaded with the mass of {sup 235}U about 1 ton. (authors)
Jiang, J.; Yuan, B.; Jin, M.; Wang, M.; Long, P.; Hu, L. [Inst. of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Univ. of Science and Technology of China, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, Univ. of Science and Technology of China, No.350 Shushanhu Road, Shushan District, Hefei, Anhui, 230031 (China)
The purpose of this paper is to document the review of several open-literature sources of both experimental capabilities and published hydrodynamic data to aid in the validation of a Computational Fluid Dynamics (CFD) based model of a slurry bubble column (SBC). The review included searching the Web of Science, ISI Proceedings, and Inspec databases, internet searches as well as other open literature sources. The goal of this study was to identify available experimental facilities and relevant data. Integral (i.e., pertaining to the SBC system), as well as fundamental (i.e., separate effects are considered), data are included in the scope of this effort. The fundamental data is needed to validate the individual mechanistic models or closure laws used in a Computational Multiphase Fluid Dynamics (CMFD) simulation of a SBC. The fundamental data is generally focused on simple geometries (i.e., flow between parallel plates or cylindrical pipes) or custom-designed tests to focus on selected interfacial phenomena. Integral data covers the operation of a SBC as a system with coupled effects. This work highlights selected experimental capabilities and data for the purpose of SBC model validation, and is not meant to be an exhaustive summary.
Donna Post Guillen; Daniel S. Wendt
Donna Post Guillen; Daniel S. Wendt; Steven P. Antal; Michael Z. Podowski
Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions
Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm lithium metaborate solution respectively at the saturation temperature for 1000 psi (68.9 bar) coolant pressure. Boiling tests also revealed the formation of fine deposits of boron and lithium on the cladding surface which degraded the heat transfer rates. The boron and lithium metaborate precipitates after a 5 day test at 5000 ppm concentration and 1000 psi (68.9 bar) operating pressure reduced the heat transfer rate 21% and 30%, respectively for the two solutions.
Schultis, J., Kenneth; Fenton, Donald, L.
We present a comprehensive study of the thermonuclear bursts and millihertz quasi-periodic oscillations (mHz QPOs) from the neutron star (NS) transient and 11 Hz X-ray pulsar IGR J17480-2446, located in the globular cluster Terzan 5. The increase in burst rate that we found during its 2010 outburst, when persistent luminosity rose from 0.1 to 0.5 times the Eddington limit, is in qualitative agreement with thermonuclear burning theory yet contrary to all previous observations of thermonuclear bursts. Thermonuclear bursts gradually evolved into a mHz QPO when the accretion rate increased, and vice versa. The mHz QPOs from IGR J17480-2446 resemble those previously observed in other accreting NSs, yet they feature lower frequencies (by a factor {approx}3) and occur when the persistent luminosity is higher (by a factor 4-25). We find four distinct bursting regimes and a steep (close to inverse cubic) decrease of the burst recurrence time with increasing persistent luminosity. We compare these findings to nuclear burning models and find evidence for a transition between the pure helium and mixed hydrogen/helium ignition regimes when the persistent luminosity was about 0.3 times the Eddington limit. We also point out important discrepancies between the observed bursts and theory, which predicts brighter and less frequent bursts, and suggest that an additional source of heat in the NS envelope is required to reconcile the observed and expected burst properties. We discuss the impact of NS magnetic field and spin on the expected nuclear burning regimes, in the context of this particular pulsar.
Linares, M.; Chakrabarty, D. [Massachusetts Institute of Technology, Kavli Institute for Astrophysics and Space Research, Cambridge, MA 02139 (United States); Altamirano, D. [Astronomical Institute 'Anton Pannekoek', University of Amsterdam and Center for High-Energy Astrophysics, P.O. BOX 94249, 1090 GE Amsterdam (Netherlands); Cumming, A. [Department of Physics, McGill University, 3600 Rue University, Montreal, QC H3A 2T8 (Canada); Keek, L. [School of Physics and Astronomy, University of Minnesota, 116 Church Street SE, Minneapolis, MN 55455 (United States)
An Experimental Examination of the Loss-of-Flow Accident Phenomenon for Prototypical ITER Divertor Channels of Y = 0 and Y = 2
This paper discusses the thermal response of two prototypical International Thermonuclear Experimental Reactor (ITER) divertor channels during simulated loss-of-flow-accident (LOFA) experiments. The thermal response was characterized by the time-to-burnout (TBO), which is a figure of merit on the mockups' survivability. Data from the LOFA experiments illustrate that (a) the pre-LOFA inlet velocity does not significantly influence the TBO, (b) the incident heat flux (IHF) does influence the TBO, and (c) a swirl tape insert significantly improves the TBO and promotes the initiation of natural circulation. This natural circulation enabled the mockup to absorb steady-state IHFs after the coolant circulation pump was disabled. Several methodologies for thermal-hydraulic modeling of the LOFA were attempted.
Marshall, Theron D. [Rensselaer Polytechnic Institute (United States); McDonald, Jimmie M. [Sandia National Laboratories (United States); Cadwallader, Lee C. [Idaho National Engineering and Environmental Laboratory (United States); Steiner, Don [Rensselaer Polytechnic Institute (United States)
An experimental examination of the loss-of-flow accident phenomenon for prototypical ITER divertor channels of Y=0 and Y=2
Marshall, T.D.; McDonald, J.M.; Cadwallader, L.C.; Steiner, D.
Fusion reactors for hydrogen production via electrolysis
The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and supplement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Depending on design, electric generation efficiencies of approx. 40 to 60% and hydrogen production efficiencies by high temperature electrolysis of approx. 50 to 70% are projected for fusion reactors using high temperature blankets.
Fillo, J A; Powell, J R; Steinberg, M
Memorandum on Chemical Reactors and Reactor Hazards
Two important problems in the investigation of reactor hazards are the chemical reactivity of various materials employed in reactor construction and the chracteristics of heat transfer under transient conditions, specifically heat transfer when driven by an exponentially increasing heat source (exp t/T). Although these problems are independent of each other, when studied in relation to reactor hazards they may occur in a closely coupled sequence. For example the onset of a dangerous chemical reactor may be due to structural failure of various reactor components under an exponentially rising heat source originating with a runaway nuclear reactor. For this reason, these two problems should eventually be studied together after an exploratory experimental survey has been made in which they are considered separately.
Mills, M.M.; Pearlman, H.; Ruebsamen, W.; Steele, G., Chrisney, J.
Thermonuclear Flame Spreading on Rapidly Spinning Neutron Stars: Indications of the Coriolis Force?
Millisecond period brightness oscillations during the intensity rise of thermonuclear X-ray bursts are likely caused by an azimuthally asymmetric, expanding burning region on the stellar surface. The time evolution of the oscillation amplitude during the intensity rise encodes information on how the thermonuclear flames spread across the stellar surface. This process depends on properties of the accreted burning layer, surface fluid motions, and the surface magnetic field structure, and thus can provide insight into these stellar properties. We present two examples of bursts from different sources that show a decrease in oscillation amplitude during the intensity rise. Using theoretical modeling, we demonstrate that the observed amplitude evolution of these bursts is not well described by a uniformly expanding circular burning region. We further show that by including in our model the salient aspects of the Coriolis force (as described by Spitkovsky, Levin, and Ushomirsky) we can qualitatively reproduce the observed evolution curves. Our modeling shows that the evolutionary structure of burst oscillation amplitude is sensitive to the nature of flame spreading, while the actual amplitude values can be very useful to constrain some source parameters.
Sudip Bhattacharyya; Tod E. Strohmayer
The TDF System for Thermonuclear Plasma Reaction Rates, Mean Energies and Two-Body Final State Particle Spectra
The rate of thermonuclear reactions in hot plasmas as a function of local plasma temperature determines the way in which thermonuclear ignition and burning proceeds in the plasma. The conventional model approach to calculating these rates is to assume that the reacting nuclei in the plasma are in Maxwellian equilibrium at some well-defined plasma temperature, over which the statistical average of the reaction rate quantity {sigma}v is calculated, where {sigma} is the cross-section for the reaction to proceed at the relative velocity v between the reacting particles. This approach is well-understood and is the basis for much nuclear fusion and astrophysical nuclear reaction rate data. The Thermonuclear Data File (TDF) system developed at the Lawrence Livermore National Laboratory (Warshaw 1991), which is the topic of this report, contains data on the Maxwellian-averaged thermonuclear reaction rates for various light nuclear reactions and the correspondingly Maxwellian-averaged energy spectra of the particles in the final state of those reactions as well. This spectral information closely models the output particle and energy distributions in a burning plasma, and therefore leads to more accurate computational treatments of thermonuclear burn, output particle energy deposition and diagnostics, in various contexts. In this report we review and derive the theoretical basis for calculating Maxwellian-averaged thermonuclear reaction rates, mean particle energies, and output particle spectral energy distributions for these reactions in the TDF system. The treatment of the kinematics is non-relativistic. The current version of the TDF system provides exit particle energy spectrum distributions for two-body final state reactions only. In a future report we will discuss and describe how output particle energy spectra for three- and four-body final states can be developed for the TDF system. We also include in this report a description of the algorithmic implementation of the TDF system, which is in two parts: the code which generates the data file itself from the nuclear reaction cross-sections, and the lookup utilities (bundled in a Fortran 77 subroutine library) which are used to retrieve the appropriate thermonuclear data from the file on demand. Unique approaches to generating, retrieving and looking up the data that were developed will be described in some detail. These developments resulted in an unusually compact data file which can be quickly generated, and from which an enormous amount of thermonuclear reaction rate, output particle energy and particle energy spectrum data can be rapidly retrieved. It is this compactness, speed and relative platform independence that distinguishes the LLNL TDF system from others.
Warshaw, S I
Laser-driven fusion reactor
A laser-driven fusion reactor consisting of concentric spherical vessels in which the thermonuclear energy is derived from a deuterium-tritium (D + T) burn within a pellet'', located at the center of the vessels and initiated by a laser pulse. The resulting alpha -particle energy and a small fraction of the neutron energy are deposited within the pellet; this pellet energy is eventually transformed into sensible heat of lithium in a condenser outside the vessels. The remaining neutron energy is dissipated in a lithium blanket, located within the concentric vessels, where the fuel ingredient, tritium, is also produced. The heat content of the blanket and of the condenser lithium is eventually transferred to a conventional thermodynamic plant where the thermal energy is converted to electrical energy in a steam Rankine cycle. (Official Gazette)
Hedstrom, J.C.
Executive Bios: Dr. Yousry Gohar - Nuclear Engineering Division...
he represented the US DOE at the International Thermonuclear Experimental Reactor (ITER) Joint Central Team in Garching (Germany), where he was the group leader for the first...
--No Title--
as project coordinator for the International Thermonuclear Experimental Reactor (ITER) engineering design activities. He became the U.S. ITER Deputy Home Team leader in...
FusEdWeb | Fusion Education
Ideas Other Fusion and Plasma Sites Great Sites Internet Plasma Physics EXperience GA's Fusion Energy Slide Show International Thermonuclear Experimental Reactor National...
House Committee on Science | Department of Energy
carrying forward with U.S. participation in the International Thermonuclear Experimental Reactor project to pursue the potential of energy from nuclear fusion. One of the biggest...
House Subcommittee on Energy and Water Development, and Related...
also are contributing to the effort known as the International Thermonuclear Experimental Reactor (ITER), to pursue the promise of clean, safe, renewable, and commercially...
??Fusion energy has been said to be the solution to all the world�s energy problems. The International Thermonuclear Experimental Reactor (ITER) is the flagship project� (more)
Mohammed, Zeshaan Sher
Astrophysics Simulations from the ASC/Alliances Center for Astrophysical Thermonuclear Flashes
The "Flash Center" works to solve the long-standing problem of thermonuclear flashes on the surfaces of compact stars such as neutron stars and white dwarfs, and in the interior of white dwarfs (i.e., Type I supernovae). The physical conditions, and many of the physical phenomena, are similar to those confronted by the Department of Energy Stockpile Stewardship program. The (fully ionized) plasmas are at very high temperatures and densities; and the physical problems of nuclear ignition, deflagration or detonation, turbulent mixing, and interface dynamics for complex multicomponent fluids are common to the weapons program. Because virtually every aspect of this problem represents a computational Grand Challenge, large-scale numerical simulations are at the heart of its resolution (Taken from Executive Summary page). More than 35 simulations and computer animations developed through research at the "Flash Center" are available here. Each .avi or .mov file also references the related research paper or presentation and provides a link.
The effects of the maximum credible accident relative to the design of the containment shell are discussed. The maximum credible accident is defined. The thermal and hydraulic effects of the maximum credible accident on the reactor system were analyzed. The extent to which fuelrod cladding will melt was estimated. The amount of energy released from the reactor system by the escaping steam and water and by a possible chemical reaction was calculated along with the corresponding pressure rise inside the containment shell. The kinds, amounts, and total radioactivity of fission products released to the atmosphere of the containment shell after the core melts were predicted. (M.C.G.)
The Conception of Thermonuclear Reactor on the Principle of Gravitational Confinement of Dense High-temperature Plasma
The work of Fisenko S. I., & Fisenko I. S. (2009). The old and new concepts of physics, 6 (4), 495, shows the key fact of the existence of gravitational radiation as a radiation of the same level as electromagnetic. The obtained results strictly correspond to the framework of relativistic theory of gravitation and quantum mechanics. The given work contributes into further elaboration of the findings considering their application to dense high-temperature plasma of multiple-charge ions. This is due to quantitative character of electron gravitational emission spectrum such that amplification of gravitational emission may take place only in multiple-charge ion high-temperature plasma.
Fisenko, Stanislav
Stanislav Fisenko; Igor Fisenko
Nuclear Reactors
Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...
Igniting the Light Elements: The Los Alamos Thermonuclear Weapon Project, 1942-1952
The American system of nuclear weapons research and development was conceived and developed not as a result of technological determinism, but by a number of individual architects who promoted the growth of this large technologically-based complex. While some of the technological artifacts of this system, such as the fission weapons used in World War II, have been the subject of many historical studies, their technical successors--fusion (or hydrogen) devices--are representative of the largely unstudied highly secret realms of nuclear weapons science and engineering. In the postwar period a small number of Los Alamos Scientific Laboratory's staff and affiliates were responsible for theoretical work on fusion weapons, yet the program was subject to both the provisions and constraints of the US Atomic Energy Commission, of which Los Alamos was a part. The Commission leadership's struggle to establish a mission for its network of laboratories, least of all to keep them operating, affected Los Alamos's leaders' decisions as to the course of weapons design and development projects. Adapting Thomas P. Hughes's ''large technological systems'' thesis, I focus on the technical, social, political, and human problems that nuclear weapons scientists faced while pursuing the thermonuclear project, demonstrating why the early American thermonuclear bomb project was an immensely complicated scientific and technological undertaking. I concentrate mainly on Los Alamos Scientific Laboratory's Theoretical, or T, Division, and its members' attempts to complete an accurate mathematical treatment of the ''Super''--the most difficult problem in physics in the postwar period--and other fusion weapon theories. Although tackling a theoretical problem, theoreticians had to address technical and engineering issues as well. I demonstrate the relative value and importance of H-bomb research over time in the postwar era to scientific, politician, and military participants in this project. I analyze how and when participants in the H-bomb project recognized both blatant and subtle problems facing the project, how scientists solved them, and the relationship this process had to official nuclear weapons policies. Consequently, I show how the practice of nuclear weapons science in the postwar period became an extremely complex, technologically-based endeavor.
Anne C. Fitzpatrick
Thermonuclear reaction rate of 18O(p,gamma)19F
For stars between 0.8-8.0 solar masses, nucleosynthesis enters its final phase during the asymptotic giant branch (AGB) stage. During this evolutionary period, grain condensation occurs in the stellar atmosphere, and the star experiences significant mass loss. The production of presolar grains can often be attributed to this unique stellar environment. A subset of presolar oxide grains features dramatic 18O depletion that cannot be explained by the standard AGB star burning stages and dredge-up models. An extra mixing process, referred to as "cool bottom processing" (CBP), was proposed for low-mass AGB stars. The 18O depletion observed within certain stellar environments and within presolar grain samples may result from the 18O+p processes during CBP. We report here on a study of the 18O(p,gamma)19F reaction at low energies. Based on our new results, we found that the resonance at Er = 95 keV (lab) has a negligible affect on the reaction rate at the temperatures associated with CBP. We also determined that the direct capture S-factor is almost a factor of 2 lower than the previously recommended value at low energies. An improved thermonuclear reaction rate for 18O(p,gamma)19F is presented.
M. Q. Buckner; C. Iliadis; J. M. Cesaratto; C. Howard; T. B. Clegg; A. E. Champagne; S. Daigle
Pinch-based thermonuclear D{sup 3}He fusion driven by a femtosecond laser
A new thermonuclear-fusion concept involving inertial-magnetic confinement on the basis of a high-current Z pinch and employing a powerful femtosecond laser is proposed. The possibility of implementing this scheme for the production of energy in a compressed liner consisting of a D{sup 3}He fuel contained inside a {sup 9}Be shell is considered. It is shown that the concerted effect of the electric discharge along the liner and an ultrashort laser shot at one of the liner endfaces provides preconditions for an ultrafast ignition of a small portion of the fuel and for the onset of the propagation of a burn wave along a plasma filament. Under favorable conditions, this system provides an energy gain of about 50, the contribution of the neutron component here being only a few percent. Basic mechanisms of the heating and confinement of the D{sup 3}He fuel are discussed along with the problem of convenient gamma-ray plasma diagnostics.
Kukulin, V. I.; Voronchev, V. T. [Moscow State University, Skobeltsyn Institute of Nuclear Physics (Russian Federation)
Review of deuterium--tritium results from the Tokamak Fusion Test Reactor
The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-{beta}{sub {ital p}} discharge following a current rampdown. The fusion power density in a core of the plasma is {approx}2.8 MW m{sup {minus}3}, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) at 1500 MW total fusion power. The energy confinement time, {tau}{sub {ital E}}, is observed to increase in D--T, relative to D plasmas, by 20% and the {ital n}{sub {ital i}}(0) {ital T}{sub {ital i}}(0) {tau}{sub {ital E}} product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-{beta}{sub {ital p}} discharges. Ion cyclotron range of frequencies (ICRF) heating of a D--T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D--T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.
McGuire, K.M.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, C.W.; Barnes, G.; Batha, S.; Bateman, G.; Beer, M.; Bell, M.G.; Bell, R.; Bitter, M.; Blanchard, W.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Cauffman, S.; Chang, Z.; Chang, C.S.; Cheng, C.Z.; Chrzanowski, J.; Collins, J.; Coward, G.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Dendy, R.; Dorland, W.; Dudek, L.; Duong, H.; Durst, R.; Efthimion, P.C.; Ernst, D.; Evenson, H.; Fisch, N.; Fisher, R.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Fujita, T.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gilbert, J.; Gioia, J.; Gorelenkov, N.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Hawryluk, R.J.; Heidbrink, W.; Herrmann, H.W.; Hill, K.W.; Hosea, J.; Hsuan, H.; Hughes, M.; Hulse, R.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kalish, M.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Labik, G.; Lam, N.T.; LaMarche, P.H.; Lawson, E.; LeBlanc, B.; Levine, J.; Levinton, F.M.; Loesser, D.; Long, D.; Loughlin, M.J.; Machuzak, J.; Majeski, R.; Mansfield, D.K.; Marmar, E.S.; Marsala, R.; Martin, A.; Martin, G.; Mazzucato, E.; Mauel, M.; McCarthy, M.P.; McChesney, J.; McCormack, B.; McCune, D.C.; McKee, G.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mirnov, S.V.; Mueller, D.; Murakami, M.; Murphy, J.A.; Nagy, A.; Navratil, G.A.; Nazikian, R.; Newman, R.; Norris, M.; O`Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D.K.; Park, H.; Park, W.; Parks, P.; Paul, S.F.; Pearson, G.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C.K.; Phillips, M.; Pitcher, S.; Pysher, R.; Qualls, A.L.; Raftopoulos, S.; Ramakrishnan, S.; Ramsey, A.; Rasmussen, D.A.; Redi, M.H.; Renda, G.; Rewoldt, G.; Roberts, D.; Rogers, J.; Rossmassler, R.; Roquemore, A.L.; Ruskov, E.; Sabbagh, S.A.; Sasao, M.; Schilling, G.; Schivell, J.; Schmidt, G.
Neutrino Oscillations with Reactor Neutrinos
Prospect measurements of neutrino oscillations with reactor neutrinos are reviewed in this document. The following items are described: neutrinos oscillations status, reactor neutrino experimental strategy, impact of uncertainties on the neutrino oscillation sensitivity and, finally, the experiments in the field. This is the synthesis of the talk delivered during the NOW2006 conference at Otranto (Italy) during September 2006.
Anatael Cabrera
Slurry reactor design studies
The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.
Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))
A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.
Treshow, M.
A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.
Daniels, F.
Thermonuclear bursts from slowly accreting neutron stars (NSs) have proven difficult to detect, yet they are potential probes of the thermal properties of the NS interior. During the first year of a systematic all-sky search for X-ray bursts using the Gamma-ray Burst Monitor aboard the Fermi Gamma-ray Space Telescope we have detected 15 thermonuclear bursts from the NS low-mass X-ray binary 4U 0614+09 when it was accreting at nearly 1% of the Eddington limit. We measured an average burst recurrence time of 12 {+-} 3 days (68% confidence interval) between 2010 March and 2011 March, classified all bursts as normal duration bursts and placed a lower limit on the recurrence time of long/intermediate bursts of 62 days (95% confidence level). We discuss how observations of thermonuclear bursts in the hard X-ray band compare to pointed soft X-ray observations and quantify such bandpass effects on measurements of burst radiated energy and duration. We put our results for 4U 0614+09 in the context of other bursters and briefly discuss the constraints on ignition models. Interestingly, we find that the burst energies in 4U 0614+09 are on average between those of normal duration bursts and those measured in long/intermediate bursts. Such a continuous distribution in burst energy provides a new observational link between normal and long/intermediate bursts. We suggest that the apparent bimodal distribution that defined normal and long/intermediate duration bursts during the last decade could be due to an observational bias toward detecting only the longest and most energetic bursts from slowly accreting NSs.
Linares, M.; Chakrabarty, D. [Massachusetts Institute of Technology, Kavli Institute for Astrophysics and Space Research, Cambridge, MA 02139 (United States); Connaughton, V.; Bhat, P. N.; Briggs, M. S.; Preece, R. [CSPAR and Physics Department, University of Alabama in Huntsville, Huntsville, AL 35899 (United States); Jenke, P.; Kouveliotou, C.; Wilson-Hodge, C. A. [Space Science Office, VP62, NASA/Marshall Space Flight Center, Huntsville, AL 35812 (United States); Van der Horst, A. J. [Astronomical Institute 'Anton Pannekoek', University of Amsterdam, NL-1090-GE Amsterdam (Netherlands); Camero-Arranz, A.; Finger, M.; Paciesas, W. S. [Universities Space Research Association, Huntsville, AL 35805 (United States); Beklen, E. [Physics Department, Suleyman Demirel University, 32260 Isparta (Turkey); Von Kienlin, A. [Max Planck Institute for Extraterrestrial Physics, Giessenbachstrasse, Postfach 1312, D-85748 Garching (Germany)
An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.
Hammond, R.P.; King, L.D.P.
Heat-transfer data are presented for the Experimental Gas Cooled Reactor Title I seven-rod fuel-assembly design. The effect on heat transfer of (1) the radial location of the outer six rods of the seven-fuel-rod cluster and of (2) the addition of helical-finned spacers at the midpoint of each of the seven fuel rods is discussed. The heattransfer data were obtained to verify preliminary general assumptions pertaining to the heat-transfer characteristics of the seven- rod fuel-assembly design and to obtain local heat-transfer correlations. The heat-transfer tests were performed at near-atmospheric pressure using air as the coolant medium. Plots and equations of heattransfer correlations over a Reynolds Number range from 12,000 to 80,000 are included. The test set-up and test procedure are also described. (auth)
Beaudoin, C.L.; Higgins, R.M.
Information on small power reactor projects of the USAEC is summarized. General information concerning the projects as a whole is given. Specific projects discussed include: the Elk River Power Reactor, the Piqua Nuclear Power Facility, the BONUS Power Reactor, the Pathfinder Power Reactor, the small-size pressurized water power reactor, and the experimental low-power process heat reactor. (M.C.G.)
Nuclear Reactor Severe Accident Experiments
Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.
Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors
The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors.
A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)
Fraas, A.P.; Mills, C.B.
National facility for advanced computational science: A sustainable path to scientific discovery
theory. Some significant problems in magnetic -confinement-fusion reactorTheory and Experiment I/O Input/output IRU Irrevocable right-of-use (networking) ISM Interstellar matter ITER International Thermonuclear Experimental Reactor (
Design, development, and testing efforts were continued on BORAX-V, EBR- I, EBR-II, EBWR, JUGGERNAUT ZPRIII, ZPR-VI, and ZPR-W. An evaluation program is outlined for Pebble Bed Reactor designs. Fast and thermal reactor safety studies were conducted. Experimental and theoretical studies in applied nuclear and reactor physics are dsscribed. Developments made in reactor components, fuels, and materials are discussed. Heat engineering studies were conducted on steam separation, and velocity and void distributions in two-phase systems. Fluidization and fluoride volatility separation, and chemical-metallurgical separation processes were studied. Advanced reactor concepts that were discusssd includsed. Basic Radiation Effects Beactor, Biogeonuclear Reactor, Fast Reactor Test Facility, compact high-power density fast reactors, AHFR hydraulic test loop, Packed Bed Reactor, and direct conversion. (For preceding period see ANL- 6328.) (B.O.G.)
A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.
Quackenbush, C.F.
Hydrolysis of CuCl{sub 2} in the Cu-Cl thermochemical cycle for hydrogen production : experimental studies using a spray reactor with an ultrasonic atomizer.
The Cu-Cl thermochemical cycle is being developed as a hydrogen production method. Prior proof-of-concept experimental work has shown that the chemistry is viable while preliminary modeling has shown that the efficiency and cost of hydrogen production have the potential to meet DOE's targets. However, the mechanisms of CuCl{sub 2} hydrolysis, an important step in the Cu-Cl cycle, are not fully understood. Although the stoichiometry of the hydrolysis reaction, 2CuCl{sub 2} + H{sub 2}O {leftrightarrow} Cu{sub 2}OCl{sub 2} + 2HCl, indicates a necessary steam-to-CuCl{sub 2} molar ratio of 0.5, a ratio as high as 23 has been typically required to obtain near 100% conversion of the CuCl{sub 2} to the desired products at atmospheric pressure. It is highly desirable to conduct this reaction with less excess steam to improve the process efficiency. Per Le Chatelier's Principle and according to the available equilibrium-based model, the needed amount of steam can be decreased by conducting the hydrolysis reaction at a reduced pressure. In the present work, the experimental setup was modified to allow CuCl{sub 2} hydrolysis in the pressure range of 0.4-1 atm. Chemical and XRD analyses of the product compositions revealed the optimal steam-to-CuCl{sub 2} molar ratio to be 20-23 at 1 atm pressure. The experiments at 0.4 atm and 0.7 atm showed that it is possible to lower the steam-to-CuCl{sub 2} molar ratio to 15, while still obtaining good yields of the desired products. An important effect of running the reaction at reduced pressure is the significant decrease of CuCl concentration in the solid products, which was not predicted by prior modeling. Possible explanations based on kinetics and residence times are suggested.
Ferrandon, M. S.; Lewis, M. A.; Alvarez, F.; Shafirovich, E.; Chemical Sciences and Engineering Division; Univ. of Texas at El Paso
Final safeguards analysis, High Temperature Lattice Test Reactor
Information on the HTLTR Reactor is presented concerning: reactor site; reactor buildings; reactor kinetics and design characteristics; experimental and test facilitles; instrumentation and control; maintenance and modification; initial tests and operations; administration and procedural safeguards; accident analysis; seifterminated excursions; main heat exchanger leak; training program outline; and reliability analysis of safety systems. (7 references) (DCC)
Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.
A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.
Wigner, E.P.
spensable tool in science, complementary to theory and experiment. It builds ...... mal experimental design of a catalytic fixed bed reactor, in Proceedings of 18th�...
Japan and France are both vying to
Experimental Reactor (ITER), a revolutionary nuclear reactor which would emulate the sun's nuclear fusion euro International Thermonuclear Experimental Reactor (ITER), a revolutionary nuclear reactor which the way for a long-awaited decision on where the nuclear reactor will be built. The United States
Attainment of Gigavolt Potentials by Fluid Dynamic Suppression of the Stepped Leader its Significance for Thermonuclear Ignition
It is proposed to levitate a conducting sphere in a high pressure Taylor flow and to charge it up to gigavolt potentials, either mechanically as in a Van de Graaff electrostatic generator, or inductively by a rising magnetic field. If the Taylor flow is sufficiently fast, it should overcome the electric pressure and breakdown by stepped leader formation, leading to the maximum attainable voltage by the Paschen law. Discharging the electrostatically stored energy can be done by controlled breakdown. With gigajoule energies stored and released in about 10^-8 sec, this implies and electric pulse power of the order 10^17 Watt, opening the prospect of large driver energies for thermonuclear ignition.
Winterberg, Friedwardt
A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.
Moore, R.V.; Bowen, J.H.; Dent, K.H.
BNL | Our History: Reactors as Research Tools
> See also: Accelerators > See also: Accelerators Brookhaven History: Using Reactors as Research Tools BGRR Brookhaven Graphite Research Reactor The Brookhaven Graphite Research Reactor (BGRR) was the Laboratory's first big machine and the first peace-time reactor built in the United States following World War II. The reactor's primary mission was to produce neutrons for scientific experimentation and to refine reactor technology. At the time, the BGRR could accommodate more simultaneous experiments than any other reactor. Scientists and engineers from every corner of the U.S. came to use the reactor, which was not only a source of neutrons for experiments, but also an excellent training facility. Researchers used the BGRR's neutrons as tools for studying atomic nuclei and the structure of solids, and to investigate many physical, chemical and
Reactor Materials
The reactor materials crosscut effort will enable the development of innovative and revolutionary materials and provide broad-based, modern materials science that will benefit all four DOE-NE...
A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.
Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.
Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.
A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)
Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.
High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.
Grebe, J.J.
Development of Applicable Benchmark Experiments for (Th,Pu)O2 Power Reactor Designs Using TSUNAMI Analysis.
?? When simulating reactor physics experiments, uncertainties in nuclear data result in a bias between simulated and experimental values. For new reactor designs or for� (more)
Langton, Stephanie E
Research reactors - an overview
A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.
West, C.D.
This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.
Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.
Brookhaven Medical Research Reactor
Medical Research Reactor BMRR The last of the Lab's reactors, the Brookhaven Medical Research Reactor (BMRR), was shut down in December 2000. The BMRR was a three megawatt...
A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)
Fortescue, P.; Nicoll, D.
A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.
Christy, R.F.
A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.
Zinn, W.H.
Catalytic reactor
A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.
Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)
A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.
Wigner, E.P.; Young, G.J.
This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors
Young, G.
A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.
Wigner, E.P.; Weinberg, A.W.; Young, G.J.
Power Burst Facility (PBF) Reactor Reactor Decommissioning
Reactor Decommissioning Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility...
LASL: controlled thermonuclear research program. Progress report, January--December 1977
Information is included for each of the following sections: (1) reversed-field pinch program, (2) Scyllac feedback stabilization experiments, (3) Scylla IV-P linear theta pinch experiments, (4) staged theta pinch, (5) field-reversal experiment, (6) implosion heating experiment, (7) fast liner experiment, (8) gun injection experiment, (9) experimental plasma physics, (10) high-density z-pinch, (11) plasma diagnostics, (12) theory, (13) computers, (14) magnetic energy transfer and storage, (15) systems studies, (16) engineering, and (17) tritium systems test assembly. (MOW)
Thomas, K.S.; Sawyer, G.A.
Advanced Safeguards Approaches for New Fast Reactors
This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to �breed� nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and �burn� actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is �fertile� or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing �TRU�-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II �EBR-II� at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line � a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.
Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.
Hydrogen production from fusion reactors coupled with high temperature electrolysis
The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and complement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Processes which may be considered for this purpose include electrolysis, thermochemical decomposition or thermochemical-electrochemical hybrid cycles. Preliminary studies at Brookhaven indicate that high temperature electrolysis has the highest potential efficiency for production of hydrogen from fusion. Depending on design electric generation efficiencies of approximately 40 to 60 percent and hydrogen production efficiencies of approximately 50 to 70 percent are projected for fusion reactors using high temperature blankets.
A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)
Neutronic reactor
A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.
Wende, Charles W. J. (West Chester, PA)
BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.
Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.
The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.
Anderson, H.L.
A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)
Post, R.G.
This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)
Starr, C.
June 25, 2005 China, South Korea oppose fusion project perks for
its bid to build an experimental nuclear fusion reactor due to opposition from China and South Korea.03 billion) reactor, was edging toward conceding its bid. A final decision on the location of the reactor known as ITER (International Thermonuclear Experimental Reactor) is expected to be made at a ministerial
MA50177: Scientific Computing Nuclear Reactor Simulation Generalised Eigenvalue Problems
MA50177: Scientific Computing Case Study Nuclear Reactor Simulation Generalised Eigenvalue of a malfunction or of an accident experimentally, the numerical simulation of nuclear reactors is of utmost balance in a nuclear reactor are the two-group neutron diffusion equations -div (K1 u1) + (a,1 + s) u1 = 1
Scheichl, Robert
The advantages, disadvantages, and cost of constructing a auclear power reactor underground are outlinedData on underground construction of hydroelectric plants, other structures, and underground reactor projects in Norway and Sweden are reviewed. A hypothetical underground Experimental Boiling Water Reactor design and sketch are given with cost estimates(T.R.H.)
Beck, C.
This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.
Leverett, M.C.
A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.
A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.
Lewis, Warren R. (Richland, WA)
An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)
Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.
Upgrading of TREAT experimental capabilities
The TREAT facility at the Argonne National Laboratory site in the Idaho National Engineering Laboratory is being upgraded to provide capabilities for fast-reactor-safety transient experiments not possible at any other experimental facility. Principal TREAT Upgrade (TU) goal is provision for 37-pin size experiments on energetics of core-disruptive accidents (CDA) in fast breeder reactor cores with moderate sodium void coefficients. this goal requires a significant enhancement of the capabilities of the TREAT facility, specifically including reactor control, hardened neutron spectrum incident on the test sample, and enlarged building. The upgraded facility will retain the capability for small-size experiments of the types currently being performed in TREAT. Reactor building and crane upgrading have been completed. TU schedules call for the components of the upgraded reactor system to be finished in 1984, including upgraded TREAT fuel and control system, and expanded coverage by the hodoscope fuel-motion diagnostics system.
Dickerman, C.E.; Rose, D.; Bhattacharyya, S.K.
This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.
Ruano, W.J.
Light Water Reactors Technology Development - Nuclear Reactors
Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy
X-10 Graphite Reactor | Department of Energy
X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor When President Roosevelt in December 1942 authorized the Manhattan Project, the Oak Ridge site in eastern Tennessee had already been obtained and plans laid for an air-cooled experimental pile, a pilot chemical separation plant, and support facilities. The X-10 Graphite Reactor, designed and built in ten months, went into operation on November 4, 1943. The X-10 used neutrons emitted in the fission of uranium-235 to convert uranium-238 into a new element, plutonium-239. The reactor consists of a huge block of graphite, measuring 24 feet on each side, surrounded by several feet of high-density concrete as a radiation shield. The block is pierced by 1,248 horizontal diamond-shaped channels in
Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression
The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.
Lasche, G.P.
Nuclear reactor
A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.
Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)
An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.
Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.
Photocatalytic reactor
A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.
Bischoff, Brian L. (Knoxville, TN); Fain, Douglas E. (Oak Ridge, TN); Stockdale, John A. D. (Knoxville, TN)
Development of an experiment to study the effects of transverse stress on the critical current of a niobium-tin superconducting cable
Superconducting magnets will play a central role for the success of the International Thermonuclear Experimental Reactor (ITER). ITER is a current driven plasma experiment that could set a milestone towards the demonstration ...
Chiesa, Luisa
The reactor components, building, control system and circuitry, and experimental and handling facilities are described and discussed, together with operation, shutdown, tank work and supplemental facilities. Training questions and answers are included. (D.C.W.)
On fast reactor kinetics studies
The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)
Seleznev, E. F.; Belov, A. A. [Nuclear Safety Inst. of the Russian Academy of Sciences IBRAE (Russian Federation); Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F. [Inst. for Physics and Power Engineering IPPE (Russian Federation)
An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.
Manley, J.H.
Nuclear Reactor Accidents
Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...
Principles of Reactor Physics
Nuclear Reactor Physics M A Smith Argonne National Laboratory Nuclear Engineering Division Phone: 630-252-9747, Email: email@example.com Abstract: Nuclear reactor physics deals with...
A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)
Accelerators for Subcritical Molten-Salt Reactors
Accelerator parameters for subcritical reactors have usually been based on using solid nuclear fuel much like that used in all operating critical reactors as well as the thorium burning accelerator-driven energy amplifier proposed by Rubbia et al. An attractive alternative reactor design that used molten salt fuel was experimentally studied at ORNL in the 1960s, where a critical molten salt reactor was successfully operated using enriched U235 or U233 tetrafluoride fuels. These experiments give confidence that an accelerator-driven subcritical molten salt reactor will work better than conventional reactors, having better efficiency due to their higher operating temperature, having the inherent safety of subcritical operation, and having constant purging of volatile radioactive elements to eliminate their accumulation and potential accidental release in dangerous amounts. Moreover, the requirements to drive a molten salt reactor can be considerably relaxed compared to a solid fuel reactor, especially regarding accelerator reliability and spallation neutron targetry, to the point that much of the required technology exists today. It is proposed that Project-X be developed into a prototype commercial machine to produce energy for the world by, for example, burning thorium in India and nuclear waste from conventional reactors in the USA.
Johnson, Roland (Muons, Inc.)
Reactor and method of operation
A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.
Wheeler, John A. (Princeton, NJ)
The agreement gives the go-ahead for work to start
international parties involved in an experimental nuclear fusion reactor project have initialed a 10bnThe agreement gives the go- ahead for work to start Fusion reactor work gets go-ahead Seven-euro (£682m) agreement on the plan. The International Thermonuclear Experimental Reactor (Iter
Precision spectroscopy with reactor anti-neutrinos
In this work we present an accurate parameterization of the anti-neutrino flux produced by the isotopes 235U, 239Pu and 241Pu in nuclear reactors. We determine the coefficients of this parameterization, as well as their covariance matrix, by performing a fit to spectra inferred from experimentally measured beta spectra. Subsequently we show that flux shape uncertainties play only a minor role in the KamLAND experiment, however, we find that future reactor neutrino experiments to measure the mixing angle $\\theta_{13}$ are sensitive to the fine details of the reactor neutrino spectra. Finally, we investigate the possibility to determine the isotopic composition in nuclear reactors through an anti-neutrino measurement. We find that with a 3 month exposure of a one ton detector the isotope fractions and the thermal reactor power can be determined at a few percent accuracy, which may open the possibility of an application for safeguard or non-proliferation objectives.
Huber, P; Huber, Patrick; Schwetz, Thomas
Accepted for publication in The Astrophysical Journal Preprint typeset using L ATEX style emulateapj v. 11/26/04 ON THE EVOLUTION OF THERMONUCLEAR FLAMES ON LARGE SCALES
The thermonuclear explosion of a massive white dwarf in a Type Ia supernova explosion is characterized by vastly disparate spatial and temporal scales. The extreme dynamic range inherent to the problem prevents the use of direct numerical simulation and forces modelers to resort to subgrid models to describe physical processes taking place on unresolved scales. We consider the evolution of a model thermonuclear flame in a constant gravitational field on a periodic domain. The gravitational acceleration is aligned with the overall direction of the flame propagation, making the flame surface subject to the Rayleigh-Taylor instability. The flame evolution is followed through an extended initial transient phase well into the steady-state regime. The properties of the evolution of flame surface are examined. We confirm the form of the governing equation of the evolution suggested by Khokhlov (1995). The mechanism of vorticity production and the interaction between vortices and the flame surface are discussed. Previously observed periodic behavior of the flame evolution is reproduced and is found to be caused by the turnover of the largest eddies. The characteristic timescales are found to be similar to the turnover time of these eddies. Relations between flame surface creation and destruction
Ju Zhang; O. E. Bronson Messer; Alexei M. Khokhlov; Tomasz Plewa
Reactor safety method
This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.
Vachon, Lawrence J. (Clairton, PA)
A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)
Ohlinger, L.A.
Experimental Neutrino Physics: Final Report
Experimental studies of neutrino properties, with particular emphasis on neutrino oscillation, mass and mixing parameters. This research was pursued by means of underground detectors for reactor anti-neutrinos, measuring the flux and energy spectra of the neutrinos. More recent investigations have been aimed and developing detector technologies for a long-baseline neutrino experiment (LBNE) using a neutrino beam from Fermilab.
Lane, Charles E.; Maricic, Jelena
An Engineering Test Reactor
A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.
Fahrner, T.; Stoker, R.L.; Thomson, A.S.
Reactor Pressure Vessel Task of Light Water Reactor Sustainability...
Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure...
X-ray and pressure conditions on the first wall of a particle beam inertial confinement reactor
Because of the presence of a chamber gas in a particle beam reactor cavity, nonneutron target debris created from thermonuclear burn will be modified or stopped before it reaches the first reactor wall. The resulting modified spectra and pulse lengths of the debris need to be calculated to determine first wall effects. Further, the cavity overpressure created by the momentum and energy exchange between the debris and gas must also be calculated to determine its effect. The purpose of this paper is to present results of the debris-background gas problem obtained with a one fluid, two temperature plasma hydrodynamic computer code model which includes multifrequency radiation transport. Spherical symmetry, ideal gas equation of state, and LTE for each radiation frequency group were assumed. The transport of debris ions was not included and all the debris energy was assumed to be in radiation. The calculated x-ray spectra and pulse lengths and the background overpressure are presented.
Magelssen, G.R.
Level 1 transient model for a molybdenum-99 producing aqueous homogeneous reactor and its applicability to the tracy reactor
Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W's proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)
Nygaard, E. T. [Babcock and Wilcox Technical Services Group, 800 Main Street, Lynchburg, VA 24504 (United States); Williams, M. M. R. [Imperial College London, SW7 2AZ (United Kingdom); Angelo, P. L. [Y-12 National Security Complex, Oak Ridge, TN 37831 (United States)
Basic and Applied Science Research Reactors - Reactors designed...
BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th...
Nuclear reactors built, being built, or planned, 1991
This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).
Simpson, B.
The purpose of this work is to present the most pertinent parts of the body of physics knowledge which has been built up in the course of the Naval and Shippingport (PWR) Reactor Programs, with the aim of providing a background of understanding for those interested in nuclear core design. Volume 1 of this handbook was planned to bring together topics in the basic theoretical and experimental material which are of especially wide interest, including those common to both thermal and intermediate neutron energy reactor types. The physics design of light water-moderated and -cooled reactors is covered in Volume 2 (classified), and that of intermediate neutron-energy power reactors in Volume 3. The emphasis in Volume 1 is thus on light water reactor systems, and as many recent advances in reactor physics of the Naval and Shippingport Reactor Programs as possible have been included.
Radkowsky, A. ed.
Validation of the RVACS (Reactor Vessel Auxiliary Cooling System)/RACS (Reactor Air Cooling System) model in SASSYS-1
The SASSYS-1 LMR systems analysis code contains a model for transient analysis of heat removal by a RVACS (Reactor Vessel Auxiliary Cooling System) or a RACS (Reactor Air Cooling System) in an LMR (Liquid Metal Reactor). This model has been validated by comparisons of model predictions with experimental data from a large scale RVACS/RACS simulation experiment performed at Argonne National Laboratory. 4 refs., 1 fig.
Dunn, F.E.
Attrition reactor system
A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.
Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)
Reactor Sharing Program
Progress achieved at the University of Florida Training Reactor (UFTR) facility through the US Department of Energy's University Reactor Sharing Program is reported for the period of 1991--1992.
Vernetson, W.G.
Guidebook to nuclear reactors
A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.
Nero, A.V. Jr.
A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.
Scott, C.D.; Davison, B.H.
A significant experimental program was conducted in the early Hanford reactors to understand the reactor production of actinides. These experiments were conducted with sufficient rigor, in some cases, to provide useful information that can be utilized today in development of benchmark experiments that may be used for the validation of present computer codes for the production of these actinides in low enriched uranium fuel.
This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)
Metcalf, H.E.
SHARP: Reactor Performance and Safety Simulation Suite
SHARP SHARP Argonne National Laboratory's Reactor Performance and Safety Simulation Suite SHARP could save millions in nuclear reactor design and development... The Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite of codes enables virtual design and engineering of nuclear plant behavior that would be impractical from a traditional experimental approach. ...by leveraging the computational power of one of the world's most powerful supercomputers. Exploiting the power of Argonne Leadership Computing Facility's near-petascale computers, researchers have developed a set of simulation tools that provide a highly detailed description of the reactor core and the nuclear plant behavior. This enables the efficient and precise design of tomorrow's safe and clean nuclear energy sources.
High solids fermentation reactor
A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.
Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)
Improved vortex reactor system
An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.
Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)
A reactor comprising fissionable material in concentration sufficiently high so that the average neutron enengy within the reactor is at least 25,000 ev is described. A natural uranium blanket surrounds the reactor, and a moderating reflector surrounds the blanket. The blanket is thick enough to substantially eliminate flow of neutrons from the reflector.
Soodak, H.; Wigner, E.P.
A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.
Epler, E.P.; Hanauer, S.H.; Oakes, L.C.
Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors
Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.
D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park
Photo of the Week: The Sixth Zero Power Reactor | Department of Energy
The Sixth Zero Power Reactor The Sixth Zero Power Reactor Photo of the Week: The Sixth Zero Power Reactor November 13, 2013 - 4:41pm Addthis In 1958, Argonne National Laboratory began the construction of several zero-power reactors (ZPRs), which are nuclear fission reactors that don't actually generate any power. Scientists developed ZPRs to assess the performance of various reactor core configurations before actually building a full nuclear reactor. A series of ZPRs were built leading up to the construction of the Experimental Breeder Reactor-II, a sodium-cooled fast reactor power plant. In this 1970 photo, an Argonne scientist is loading the matrices of the ZPR-VI reactor prior to its first operation using plutonium fuel. | Photo courtesy of the Department of Energy. In 1958, Argonne National Laboratory began the construction of several
print close Mon 2 May 2005
nuclear fusion reactor. The two sides have been at pains to craft a compromise over competing Japaneseprint close Mon 2 May 2005 (UK)2:56am Europe and Japan at Odds over 'Superpower' Reactor "PA and French bids to host the £6.8billion International Thermonuclear Experimental Reactor. The ITER plant
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means "the way" in Latin, is based on the idea. All the commercialized nuclear reactors in the world's winning to host nuclear fusion reactor France wins bid to host nuclear fusion project Chirac delighted to build its own thermonuclear experimental reactor A leading Chinese plasma physicist said Thursday China
Wednesday, February 4, 2004 Search Archives
, repeated Japan's desire to host the world's first prototype nuclear fusion reactor in Rokkasho, Aomori to iron out their differences over the site of the International Thermonuclear Experimental Reactor. Russia sticks to its guns, backs France for fusion reactor Yamato Inaba, senior vice science minister
Sunday, June 20, 2004 SPORTS LIFE
nations failed Friday to forge an accord on whether to build the world's first prototype nuclear fusion reactor in France or Japan. Decision on site for fusion reactor put off again In a meeting of sub-Cabinet officials from the six parties involved in the International Thermonuclear Experimental Reactor project
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to host test reactor in search for alternative to fossil fuel French nuclear research facilities near Tuesday. The International Thermonuclear Experimental Reactor, or ITER, is intended to show that nuclear, and the creation of thousands of jobs. Fusion, not fission Unlike fission reactors used in existing nuclear power
Greenpeace International Jun 28, 2005 @ 11:06 am
archive Posted on: Nuclear fusion reactor project in France: an expensive and senseless nuclear stupidity to the International Thermonuclear Experimental Reactor (ITER) (1) to construct one of the world's largest nuclear reactor project in France: an expensive and senseless nuclear stupidity Page 1 of 1http
The Argonaut Reactor - Reactors designed/built by Argonne National
Achievements > Achievements > Argonne Reactors > Training Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy
Nuclear reactor overflow line
The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.
Severson, Wayne J. (Pittsburgh, PA)
Reactor vessel support system
A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.
Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)
Japanese set to direct `sun-power' nuclear reactor in France September 16, 2005
Japanese set to direct `sun-power' nuclear reactor in France September 16, 2005 Japan has been develop three generations of nuclear reactors and includes six low-capacity experimental reactors and a 17 asked to nominate the chief of an international project to build a multi- billion-dollar nuclear fusion
Spinning fluids reactor
A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.
Miller, Jan D; Hupka, Jan; Aranowski, Robert
Determining Reactor Neutrino Flux
Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.
Cao, Jun
Reactor water cleanup system
A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.
Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)
Jun Cao
Fission reactors and materials
The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.
Frost, B.R.T.
Role of research reactors in training of NPP personnel with special focus on training reactor VR-1
Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training program are demonstrated. (authors)
Sklenka, L.; Rataj, J.; Frybort, J.; Huml, O. [Dept. of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical Univ. in Prague, V Holesovickach 2, Prague 8, 180 00 (Czech Republic)
Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment |
Obtains Patent for Nuclear Reactor Sodium Cleanup Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment March 28, 2013 - 12:00pm Addthis CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. Piping in the east boiler basement of the sodium processing building was color coded for easy identification. Orange indicates sodium and green identifies cooling water.
Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment Idaho Site Obtains Patent for Nuclear Reactor Sodium Cleanup Treatment March 28, 2013 - 12:00pm Addthis CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. CWI engineers Jeff Jones, David Tolman, right, and Kirk Dooley (seated) developed a treatment to safely dissolve a bicarbonate crust and treat and remove the sodium in the Experimental Breeder Reactor-II at the Idaho site. Piping in the east boiler basement of the sodium processing building was color coded for easy identification. Orange indicates sodium and green identifies cooling water.
Two days of negotiations in
-dollar experimental nuclear reactor amid an intense contest with France and is ready to discuss ways to appease France or Japan would host the revolutionary International Thermonuclear Experimental Reactor (ITER) project. BACK JAPAN DETERMINED TO HOST NUCLEAR PROJECT, READY TO DISCUSS COMPENSATION Received Thursday
Board on Physics and Astronomy 500 Fifth Street, NW, Washington, D.C. 20001
--is a key fusion nuclear technology. Fusion reactors must operate with more tritium produced and recovered on the International Thermonuclear Experimental Reactor (ITER), an international burning plasma experiment.1 in understanding and predicting plasma performance-- particularly in the field of plasma theory and experimental
Generation -IV Reactor Concepts
Generation-IV Reactor Concepts Generation-IV Reactor Concepts Thomas H. Fanning Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439, USA The Generation-IV International Forum (GIF) is a multi-national research and development (R&D) collaboration. The GIF pursues the development of advanced, next generation reactor technology with goals to improve: a) sustainability (effective fuel utilization and minimization of waste) b) economics (competitiveness with respect to other energy sources) c) safety and reliability (e.g., no need for offsite emergency response), and d) proliferation resistance and physical protection The GIF Technology Roadmap exercise selected six generic systems for further study: the Gas- cooled Fast Reactor (GFR), the Lead-cooled Fast Reactor (LFR), the Molten Salt Reactor (MSR),
Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.
A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.
Spinrad, B.I.
An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.
Diebold, J.P.; Scahill, J.W.
A book is presented which supplies pertinent seismological information to engineers in the nuclear reactor field. Data are presented on the occurrence, intensity, and wave shapes. Techniques are described for evaluating the response of structures to such events. Certain reactor types and their modes of operation are described briefly. Various protection systems are considered. Earthquake experience in industrial and reactor plants is described. (D.L.C.)
This report describes the past procedure in dismantling the reactor block of the FRJ-1 research reactor (MERLIN). Furthermore, it gives an outlook on future activities up to the final removal of the reactor block. MERLIN is an abbreviation for Medium Energy Research Light Water Moderated Industrial Nuclear Reactor. The FRJ-1 (MERLIN) was shut down in 1985 and the fuel elements removed from the facility. After dismantling the coolant loops and removing the reactor tank internals with subsequent draining of the reactor tank water, the first activities for dismantling the reactor block were carried out in summer 2001. The relevant license was granted in late July 2001 by the licensing authority specifying 8 incidental provisions. After dismantling the reactor extension (gates of the thermal columns and steel platforms surrounding the reactor block), a heavy-load platform including a casing around the reactor block was constructed. Two ventilation systems with a volume flow of 10,000 and 2 ,000 m3/h will, moreover, serve to avoid a spread of contamination. The reactor block will be dismantled in three phases divided according to upper, central and bottom sections. Dismantling the upper section started in August 2002. This section as well as the bottom section can probably be completely measured for clearance. For this reason, the activities have so far been carried out manually using mechanical and thermal techniques. The central section will probably have to be largely disposed of as radioactive waste. This is the region of the former reactor core in which the experimental devices are also integrated. Most of this work will probably have to be carried out by remote handling. More than 80 % of the dismantled materials of the reactor block can probably be measured for clearance. For this purpose, a clearance measurement device was taken into operation in the FRJ-1. On this occasion, the limits of clearance measurement have become evident. For concrete, which constitutes the largest portion of the dismantled materials by volume, an additional conditioning step has become necessary to fulfill the clearance criteria, whereas waste packages with steel components largely have to be reconditioned once more at a later stage. Material measured for clearance will be disposed of conventionally (recycling, landfill) after inspection by the official expert and clearance by the regulatory authority. Dismantled parts that cannot be measured for clearance will be transferred to the Decontamination Department of the Research Centre. From the present perspective, the dismantling of the reactor block will be completed within the first six months of 2003.
Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.
Pressurized fluidized bed reactor
A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.
Isaksson, Juhani (Karhula, FI)
A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.
Isaksson, J.
Tokamak reactor first wall
This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.
Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.
Advanced Nuclear Research Reactor
This report describes technical modifications implemented by INVAP to improve the safety of the Research Reactors the company designs and builds.
Lolich, J.V.
A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.
King, L.D.P.
Status of LDRD-DR 20070518 development of a magnetically driven target for thermo-nuclear burn studies (u)
This project is developing a magnetically driven cylindrical confinement system for the creation of a small region of material existing under extreme conditions. Using a Ranchero High Explosive Pulsed Power generator (HEPP) with maximum current ranging from 25- 50 MA depending on the load, a current driven Al cylinder will impact a series of nested, less massive Au shells. Each subsequent shell's inner surface velocity will increase due to it's smaller mass by the ratio 2.01( 1+ m{sub i+ 1}/m i), along with radial convergence. Attaining this ideal result requires highly efficient energy transfer which in turn requires plastic cushions. The final velocity of the last sequential shell will be used to drive a central experimental package in which extreme material conditions will be produced. The inexpensive nature of HEPP and the extreme conditions attainable allow many studies to be conducted in regimes not currently available in the laboratory. One potential central experimental package consists of a cylindrical Inertial Confinement Fusion (ICF) target; a cylindrical Au pusher surrounding frozen DT. This target is used as a design tool. The ICF conditions achieved with such a target would be similar to those created in a double shell ignition capsule at the National Ignition Facility. The system being developed has a range of potential applications.
Watt, Robert G [Los Alamos National Laboratory; Atchison, W L [Los Alamos National Laboratory; Colgate, S A [Los Alamos National Laboratory; Gofoeth, J [Los Alamos National Laboratory; Griego, J [Los Alamos National Laboratory; Guzik, J [Los Alamos National Laboratory; Holtkamp, D [Los Alamos National Laboratory; Idzorek, G [Los Alamos National Laboratory; Kirkpatrick, R [Los Alamos National Laboratory; Menikoff, R [Los Alamos National Laboratory; Meyer, R [Los Alamos National Laboratory; Oona, H [Los Alamos National Laboratory; Reardon, P [Los Alamos National Laboratory; Rousculp, C L [Los Alamos National Laboratory; Sgro, A G [Los Alamos National Laboratory; Tabaka, L [Los Alamos National Laboratory
Experimental design
Maximizing data information requires careful selection, termed design, of the points at which data are observed. Experimental design is reviewed here for broad classes of data collection and analysis problems, including: fractioning techniques ... Keywords: Key Design Issues in Data Mining, Science and Technology, Statistical Fundamentals
J. P. Morgan; Xinwei Deng
Foreign Research Reactor/Domestic Research Reactor Receipt Coordinator...
Research ReactorDomestic Research Reactor Receipt Coordinator, Savannah River Nuclear Solutions | National Nuclear Security Administration Our Mission Managing the Stockpile...
Organization of the ORNL Operations division for administration of the Oak Ridge Research Reactor, the Low Intensity Testing Reactor, and the Oak Ridge Graphite Reactor is described. (J.R.D.)
Casto, W.R.
Production reactor characteristics
Reactors for the production of special nuclear materials share many similarities with commercial nuclear power plants. Each relies on nuclear fission, uses uranium fuel, and produces large quantities of thermal power. However, there are some important differences in production reactor characteristics that may best be discussed in terms of mission, role, and technology.
Thiessen, C.W.; Hootman, H.E.
Advanced converter reactors
Advanced converter reactors (ACRs) of primary US interest are those which can be commercialized within about 20 years, and are: Advanced Light-Water Reactors, Spectral-Shift-Control Reactors, Heavy-Water Reactors (CANDU type), and High-Temperature Gas-Cooled Reactors. These reactors can operate on uranium, thorium, or uranium-thorium fuel cycles, but have the greatest fuel utilization on thorium type cycles. The water reactors tend to operate more economically on uranium cycles, while the HTGR is more economical on thorium cycles. Thus, the HTGR had the greatest practical potential for improving fuel utilization. If the US has 3.4 to 4 million tons U/sub 3/O/sub 8/ at reasonable costs, ACRs can make important contributions to maintaining a high nuclear power level for many decades; further, they work well with fast breeder reactors in the long term under symbiotic fueling conditions. Primary nuclear data needs of ACRs are integral measurements of reactivity coefficients and resonance absorption integrals.
Kasten, P.R.
A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.
A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.
The Integral Fast Reactor
Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.
Till, C.E.; Chang, Y.I. (Argonne National Lab., IL (USA)); Lineberry, M.J. (Argonne National Lab., Idaho Falls, ID (USA))
Computer simulations help design new nuclear reactors
Computer simulations help design new nuclear reactors Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share Reprinted from "Argonne Now" - Spring 2008 Physicist Won-Sik Yang and computer scientist Andrew Siegel hold a fuel rod assembly in front of a model of the Experimental Breeder Reactor-II
GAS COOLED NUCLEAR REACTOR STUDY. Final Report
An investigntion was made of the performance of a gas-cooled reactor, designed to provide a source of high temperature heat to a stream of helium. This reactor, in turn, is used as a source of heat for the air stream in a gas- turbine power plant. The reactor design was predicted primarily on the requirement for transferring a large amount of heat to the helium stream with a pressure drop low enough that it will not represent a major loss of power in the power plant. The mass of uranium e uired far criticality under various circumstances was investigated by multigroup calculations, both on desk calculators and on an IBM-704 machine. The gasturbine power plant perfarmance was studied based on a Studebaker-Packard-designed gas-turbine power plant for the propulsion of destroyer-escort vessels. A small experimental program was carried out to study some effects of helium on graphite and on structural steels. (auth)
Thompson, A.S.
improved Bevalac Fusion reactors Vi Q) U .c. Laser, expl.into thermonuclear fusion reactors. A summary of ion sources
Clark, D.J.
Progress on reactor programs and in general engineering research and development programs is summarized. Research and development are reported on water-cooled reactors including EBWR and Borax-V, sodium-cooled reactors including ZPR-III, IV, and IX, Juggernaut, and EBR-I and II. Other work included a review of fast reactor technology, and studies on nuclear superheat, thermal and fast reactor safety, and reactor physics. Effort was also devoted to reactor materials and fuels development, heat engineering, separation processes and advanced reactor concepts. (J.R.D.)
Nuclear reactor control column
The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.
Bachovchin, Dennis M. (Plum Borough, PA)
GCFR shielding design and supporting experimental programs
The shielding for the conceptual design of the gas-cooled fast breeder reactor (GCFR) is described, and the component exposure design criteria which determine the shield design are presented. The experimental programs for validating the GCFR shielding design methods and data (which have been in existence since 1976) are also discussed.
Perkins, R.G.; Hamilton, C.J.; Bartine, D.
Nuclear reactor reflector
A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.
Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)
A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.
Hopkins, R.J.; Land, J.T.; Misvel, M.C.
Microfluidic electrochemical reactors
A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.
Nuzzo, Ralph G. (Champaign, IL); Mitrovski, Svetlana M. (Urbana, IL)
Fast Breeder Reactor studies
This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.
Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.
Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.
Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.
Spherical torus fusion reactor
The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.
Martin Peng, Y.K.M.
S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.
Lichtenberger, H.V.; Cameron, R.A.
Design Considerations for Economically Competitive Sodium Cooled Fast Reactors
The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Ph�nix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.
Hongbin Zhang; Haihua Zhao
Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel
The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.
Hayes, A C
A. C. Hayes; Gerard Jungman
The design and construction program of the Bolling Experimental Reactor is reviewed. A number of preliminary experinnents were performed with Borax-II at pressures between 75 and 300 psi. The most corrosion-resistant U-Zr--Nb alloy developed so far is produced by heating in a vacuum to the gamma phase, quenching, and aging for 2 hr at 400 deg C. Special attention is given to the removal of H/sub 2/ from the material. Unclad and unirradiated samples of U--Nb and U--Nb--Zr alloys were corrosion tested in H/sub 2/O. Corrosion rates were also measured under irradiation conditions in CP-5. Elongation measurements of irradiated wrought and cast U--Zr material suggested no way for treating the wrought fuel so that stability comparable to the cast material could be obtained. Natural circulation boiling density tests at 600 psia were made in order to determine the effects of channel cross section and subcooling on the steam void fraction. Results of autoclave and dynamic corrosion studies of 2-S Al in H/sub 2/O are reported. These results include the testing of Ni-clad samples. A large number of criticality calculations were performed for the EBR-Il and the PBR. The solubility of ThO/sub 2/ pellets containing various concentrations of U/sub 3/ O/sub 8/ was tested in water at 316 deg C for periods of 672 to 744 hr. None of the samples disintegrated, although at least one sample developed cracks. Solutions of reactor kinetic equations were attempted for the purpose of studying transients in reactors with lifetimes of 7 x 10/sup -9/ 10/sup -7/, and 6 x 10/ sup -5/ sec. Ignition experiments were performed on Th, Cu, Al, Fe, Mg, Zr, and fluorothene when contacted with fluorides. Except for Zr and fluorothene, the materials did not ignite. (C.H.)
Reactor hot spot analysis
The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.
Vilim, R.B.
Reactor Thermal-Hydraulics
Thermal-Hydraulics Thermal-Hydraulics Dr. Tanju Sofu, Argonne National Laboratory In a power reactor, the energy produced in fission reaction manifests itself as heat to be removed by a coolant and utilized in a thermodynamic energy conversion cycle to produce electricity. A simplified schematic of a typical nuclear power plant is shown in the diagram below. Primary coolant loop Steam Reactor Heat exchanger Primary pump Secondary pump Condenser Turbine Water Although this process is essentially the same as in any other steam plant configuration, the power density in a nuclear reactor core is typically four orders of magnitude higher than a fossil fueled plant and therefore it poses significant heat transfer challenges. Maximum power that can be obtained from a nuclear reactor is often limited by the
Integral physics data for fast reactor design are discussed. The measurements needed include those of critical mass, shape factor, detector ratios, neutron spectra, material replacement experiments, reflector savings, neutron lifetime, Rossi- alpha , and similar quantities. Topics covered include Pu- and U/sup 233/-fueled systems, highly enriched U/sup 235/ systems in optimum geometry, uranium cores of various enrichments and dilutions, extreme geometry critical experiments, specific reactor systems, core mockup inhomogeneities, spectral studies and detector ratios, uranium equilibrium spectrum data, materialreplacement measurements, fast reactor dynamics, and suggested future experiments and experimental programs. (M.C.G.)
Loewenstein, W.B.; Meneghetti, D.
Compact power reactor
There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.
Wetch, Joseph R. (Woodland Hills, CA); Dieckamp, Herman M. (Canoga Park, CA); Wilson, Lewis A. (Canoga Park, CA)
Molten metal reactors
A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.
Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M
Reactor Safety Research Programs Quarterly Report July - September 1981
This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.
Edler, S. K.
System and method for generating steady state confining current for a toroidal plasma fusion reactor
A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma.
Fisch, Nathaniel J. (Cambridge, MA)
A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma.
Bers, Abraham (Arlington, MA)
A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.
West, J.M.; Weills, J.T.
Detailed search Miscellaneous
; Nuclear Fusion France and Japan disagree over outcome of ITER meeting [Date: 2005-05-09] Discussions over the eventual site of the international thermonuclear experimental reactor (ITER) have become more heated after. 'This well-balanced agreement could lead Japan to decide not to be the host country for the reactor
By Dr David Whitehouse BBC News Online science editor
The International Thermonuclear Experimental Reactor (Iter) is the boldest nuclear initiative since the Manhattan - NUCLEAR FUSION PROJECT The project is estimated to cost $5bn over the next 10 years It will produce the first sustained fusion reactions Iter is the final stage before a commercial reactor is built Scientists
Koizumi stands firm to Chirac on Japan hosting nuclear project Prime Minister Junichiro Koizumi said Sunday that Japan stood firm in its attempt to host a
Koizumi said Sunday that Japan stood firm in its attempt to host a revolutionary nuclear reactor, amid EUKoizumi stands firm to Chirac on Japan hosting nuclear project 27/03/2005 Prime Minister Junichiro Jacques Chirac discussed the deadlock over the International Thermonuclear Experimental Reactor (ITER
Monday, February 23, 2004 Decision on site for fusion project is put off again
the world's first prototype nuclear fusion reactor, according to Satoru Otake of the science and technology) The six parties involved in an international nuclear fusion project have again failed to decide on either in March. Senior officials of the parties to the International Thermonuclear Experimental Reactor project
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together, as opposed to today's nuclear reactors and weapons, which produce energy by blowing atoms apart nuclear fusion program, a ministry spokesman said Tuesday. Japan Minister Plans Push for Nuclear Fusion, both of which are bidding for the International Thermonuclear Experimental Reactor (ITER). The European
Friday February 20, 2004 Three compromise plans eyed for fusion project site
unable to agree where to locate the world's first prototype nuclear fusion reactor. The proposals come on an international nuclear fusion project has drawn up three compromise proposals in a bid to resolve the row over said. The six parties involved in the Thermonuclear Experimental Reactor (ITER) project have been
France to win huge nuclear fusion project 10:42 24 June 2005
Edwards A long and bitter dispute about where to site the world's largest nuclear fusion reactor looks allFrance to win huge nuclear fusion project 10:42 24 June 2005 NewScientist.com news service Rob the International Thermonuclear Experimental Reactor (ITER) at Rokkashomura in Japan or at Cadarache in France
-News Home Help EU presses Japan to compromise on nuclear project
-News Home Help EU presses Japan to compromise on nuclear project 2 hours, 23 minutes ago BRUSSELS to resolve a standoff over who will host a revolutionary nuclear reactor project, a key EU official said Thermonuclear Experimental Reactor (ITER), which has been billed as a test bed for a safe and inexhaustible
Experimental Boiling Water Reactor decontamination and decommissioning project
The author begins by discussing the problems encountered during decontamination and decommissioning. Next, he discusses waste packaging and recycling. His last topic of lessons learned is subdivided into prevention and early detection, recovery issues, management issues, and noteworthy practices.
Fellhauer, C. [Argonne National Lab., IL (United States). Technology Development Div.
Magnetic shielding system for the tokamak experimental power reactor
A poloidal field system is proposed for superconducting tokamaks which magnetically shields the toroidal field coil windings from pulsed fields and reduces the plasma inductive requirements. The principal elements of the system are a primary air core winding and a shield-vertical-field winding. The properties of the system are investigated using computer codes for MHD equilibrium, inductances, and fields. (auth)
Marcus, F.B.; Peng, Y.K.M.; Dory, R.A.; Moore, J.R.
Nuclear reactors built, being built, or planned: 1995
This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).
Nuclear reactors built, being built, or planned, 1994
This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).
Sodium fast reactor safety and licensing research plan. Volume II.
Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.
Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d'%C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.
Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.
Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.
Progress is reported on investigations in support of the Experimental Gas-Cooled Reactor, the Pebble-Bed Reactor Experiment, Advanced reactor design and development, test facilities, components, and materials. Topics covered include EGCR physics, EGCR performance analyses, structural investigations, EGCR component and materials development and testing, EGCR experimental facilities, PBRE physics and design studies, fueled-graphite investigations, clad fuel development, design studies of advanced power plants, experimental investigations of heat transfer and fluid flow, development of equipment anmd test facilities. and fabrication studies. (M.C.G.)
B Reactor | Department of Energy
Operational Management » History » Manhattan Project » Signature Operational Management » History » Manhattan Project » Signature Facilities » B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first large-scale plutonium production reactor. As at Oak Ridge, the need for labor turned Hanford into an atomic boomtown, with the population reaching 50,000 by summer 1944. Similar to the X-10 Graphite Reactor at Oak Ridge in terms of loading and unloading fuel, the B Reactor was built on a much larger scale and used water rather than air as a coolant. Whereas the X-10 had an initial design output of 1,000 kilowatts, the B Reactor was designed to operate at 250,000 kilowatts. Consisting of a 28- by 36-foot, 1,200-ton graphite cylinder lying on its side, the reactor was penetrated through its
Recycle of LWR (Light Water Reactor) actinides to an IFR (Integral Fast Reactor)
A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs.
Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.
Gas-cooled reactors
Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing.
Schulten, R.; Trauger, D.B.
Zero Power Reactor simulation | Argonne National Laboratory
Zero Power Reactor simulation Share Description Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by...
jet aircraft engines, and nuclear reactor fuel elements. Ancomponents of a nuclear reactor core are susceptible tothe nuclear physics of the thermal and fast neutron reactors
Olander, Donald R.
Advanced Reactor Development and Technology - Nuclear Engineering...
Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Nuclear Data Program Advanced Reactor Development Overview Advanced Fast Reactor...
Development of high temperature catalytic membrane reactors
Significant progress was made in 1991 on the development of ceramic membranes as catalytic reactors. Efforts were focused on the design, construction and startup of a reactor system capable of duplicating relevant commercial operating conditions. With this system, yield enhancement was demonstrated for the dehydrogenation of ethylbenzene to styrene in a membrane reactor compared to the standard packed bed configuration. This enhancement came with no loss in styrene selectivity. During operation, coke deposition on the membrane was observed, but this deposition was mitigated by the presence of steam in the reaction mixture and a steady state permeability was achieved for run times in excess of 200 hours. Work began on optimizing the membrane reactor by exploring several parameters including the effect of N{sub 2} diluent in the reaction feed and the effect of a N{sub 2} purge on the permeate side of the membrane. This report details the experimental progress made in 1991. Interactions with the University of Wisconsin on this project are also summarized. Finally, current status of the project and next steps are outlined.
Nuclear reactor safety device
A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.
Hutter, Ernest (Wilmette, IL)
A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)
Graham, R.H.
A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.
Luebke, E.A.; Vandenberg, L.B.
A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.
Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.
Heat dissipating nuclear reactor
Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.
Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)
Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.
Hunsbedt, A.; Lazarus, J.D.
Reactor for exothermic reactions
A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.
Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.
A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)
Sankovich, M.F.; Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Gestson, D.K.
Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:
Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from
Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including
Nuclear reactor apparatus
A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.
Wade, Elman E. (Ruffs Dale, PA)
Fusion reactor pumped laser
A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.
Jassby, Daniel L. (Princeton, NJ)
Perspectives on reactor safety
The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.
Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)
Reactor Neutrino Experiments
Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measurement are also briefed.
Diagnostics for hybrid reactors
The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.
Orsitto, Francesco Paolo [ENEA Unita' Tecnica Fusione , Associazione ENEA-EURATOM sulla Fusione C R Frascati v E Fermi 45 00044 Frascati (Italy)
Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.
Fenning, F.W.; Jackson, R.F.
A reactor of the heterageneous, heavy water moderated type is described. The reactor is comprised of a plurality of vertically disposed fuel element tubes extending through a tank of heavy water moderator and adapted to accommodate a flow of coolant water in contact with the fuel elements. A tank containing outgoing coolant water is disposed above the core to function is a radiation shield. Unsaturated liquid hydrocarbon is floated on top of the water in the shield tank to reduce to a minimum the possibility of the occurrence of explosive gaseous mixtures resulting from the neutron bombardment of the water in the shield tank.
Garrison, W.M.; McClinton, L.T.; Burton, M.
Fast quench reactor method
A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.
Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)
Reactor Safety Research Programs Quarterly Report October - December 1980
This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.
Edler, S K
Reactor sharing experience at the MIT research reactor
This paper provides a number of examples of how educational institutions in the Boston area and elsewhere that do not possess nuclear reactors for training and research purposes have successfully enriched their programs through utilization of the Massachusetts Institute of Technology Research Reactor (MITR) with assistance from the Reactor Sharing Program of the US Department of Energy (DOE).
Clark, L. Jr.; Fecych, W.; Young, H.H.
The considerations which led to the study of a homogeneous suspension reactor are reviewed briefly. The characteristics of the KEMA Suspension Test Reactor (KSTR) are then summarized. (J.S.R.)
Went, J.J.
Waste tyre pyrolysis in a conical spouted bed reactor under vacuum conditions
Pyrolysis of scrap tires has been studied at 425 and 500�C and three pressure levels, atmospheric one 0.5 and 0.25 atm. The experimental runs have been carried out in a pilot plant conical spouted bed reactor provided with a system for continuous ... Keywords: carbon black, conical spouted bed reactor, tire pyrolysis, vacuum pyrolysis
Gartzen Lopez; Maider Amutio; Gorka Elordi; Maite Artetxe; Aitziber Erkiaga; Astrid Barona; Martin Olazar
Advanced Nuclear Reactors | Department of Energy
in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor...
Innovative design of uranium startup fast reactors
Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...
Fei, Tingzhou
The results of a study are presented which provided additional experimental-loop irradiation space for the AECDRD testing program. It was a premise that the experiments allocated to this reactor were those which could not be accommodated in the MTR, ETR, or in existing commercial test reactors. To accomplish the design objectives called for a reactor producing perturbed neutron fluxes exceeding 1O/sup 15/ thermal n/cm/sup 2/-sec and 1.5 x 1O/sup 15/ epithermal n/cm/sup 2/-sec. To accommodate the experimental samples, the reactor fuel core is four feet long in the direction of experimental loops. This is twice the length of the MTR core and a third longer than the ETR core. The vertical arrangement of reactor and experiments permits the use of loops penetrating the top cap of the reactor vessel running straight and vertically through the reactor core. The design offers a high degree of accessibility of the exterior portions of the experiments and offers very convenient handling and discharge of experiments. Since the loops are to be integrated into the reactor design and the in-pile portions installed before reactor start-up, it is felt that many of the problems encountered in MTR and ETR experience will cease to exist. Installation of the loops prior to startup will have an added advantage in that the flux variations experienced in experiments in ETR every time a new loop is installed will be absent. The Advanced Test Reactor has a core configuration that provides essentially nine flux-trap regions in a geometry that is almost optimum for cylindrical experiments. The geometry is similar to that of a fourleaf clover with one flux trap in each leaf, one at the intersection of the leaves, and one between each pair of leaves. The nominal power level is 250 Mw. The study was carried out in enough detail to permit the establishment of the design parameters and to develop the power requirement which, conservatively rated, will definitely reach the flux specifications. A critical mockup of an arrangement similar to ATR was loaded into the Engineering Test Reactor Critical Facility. (auth)
deBoisblanc, D.R. et al
Reactor operation safety information document
The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)
A reactor fuel element of the capillary tube type is described. The element consists of a thin walled tube, sealed at both ends, and having an interior coatlng of a fissionable material, such as uranium enriched in U-235. The tube wall is gas tight and is constructed of titanium, zirconium, or molybdenum.
Kesselring, K.A.; Seybolt, A.U.
A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.
Horning, W.A.; Lanning, D.D.; Donahue, D.J.
As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.
The First Reactor
Chicago Pile-1�(CP-1) was the world's first�nuclear reactor.�CP-1 was built on a�rackets�court, under the abandoned west stands of the original�Alonzo Stagg Field�stadium, at the�University of...
Thermal Reactor Safety
Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.
A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)
Cobb, W.G.
Cermet fuel reactors
Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.
Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.
Nuclear reactor building
A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.
Gou, P.F.; Townsend, H.E.; Barbanti, G.
Neutronic reactor thermal shield
1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.
Fossil fuel furnace reactor
A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.
Parkinson, William J. (Los Alamos, NM)
Nuclear Reactors and Technology
This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.
Cason, D.L.; Hicks, S.C. [eds.
A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.
Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)
Reactor component automatic grapple
A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.
Greenaway, Paul R. (Bethel Park, PA)
A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)
Currier, E.L. Jr.; Nicklas, J.H.
A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.
Weinberg, A.M.; Vernon, H.C.
The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)
Vernon, H.C.; Weinberg, A.M.
The reactor radiation shield material is comprised of alternate layers of iron-containing material and compressed cellulosic material, such as masonite. The shielding material may be prefabricated in the form of blocks, which can be stacked together in ary desired fashion to form an effective shield.
Fermi, E.; Zinn, W.H.
A boron-10 containing reactor control element wherein the boron-10 is dispersed in a matrix material is describeri. The concentration of boron-10 in the matrix varies transversely across the element from a minimum at the surface to a maximum at the center of the element, prior to exposure to neutrons. (AEC)
Beaver, R.J.; Leitten, C.F. Jr.
The three main chemical problems of non-aqueous fluidfuel reactors are selection of a fuel system which meets nuclear and thermal requirements, control of corrosion of structural materials, and development of an efficient and economical separation process. The problems of the fastneutron reactor with a core of fused NaCl, PbCl/sub 2/, UCl/sub 4/ mixture and a blanket of fused UCl/ sub 4/ are discussed, except for the separation process. Brief treatment is given the Bi- U - Pu fuel system for thermal reactors, including tentative flowsheets for the separation process. A detailed discussion of the experiments of Bareis at Brookhaven and related experiments on the distribution of various metals between liquid Bi and fused-salt solutions is given, the experimental results correlated, and application made to reactor problems. A general discussion is given of nonaqueous high-temperature separation processes (L.M.T.)
Scatchard, G.; Clark, H.M.; Golden, S.; Boltax, A.; Schuhmann, R. Jr.
This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)
Whitham, G.K.; Smith, R.R.
An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)
Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.
Modeling Chemical Reactors I: Quiescent Reactors
We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(...
Michoski, C E; Schmitz, P G
Static corrosion of candidate alloys for the lead-bismuth fast reactor
This project examined the corrosion rates and mechanisms of two candidate alloys for use in Lead-bismuth Eutectic (LBE) cooled fast nuclear reactors. The two alloys examined were T91 and Fe-12Cr-2Si. An experimental study ...
Wang, Yunzhi (Yunzhi Diana)
Survey of Optimization of Reactor Coolant Cleanup Systems: For Boiling Water Reactors and Pressurized Water Reactors
Optimization of the reactor coolant cleanup systems in the boiling water reactor (BWR) and pressurized water reactor (PWR) environment is important for controlling the transport of corrosion products (metals and activated metals), fission products, and coolant impurities (soluble and insoluble) throughout the reactor coolant loop, and this optimization contributes to reducing primary system radiation fields. The removal of radionuclides and corrosion products is just one of many functions (both ...
Advanced burner test reactor preconceptual design report.
The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.
Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA
Laser decontamination and cleaning of metal surfaces : modelling and experimental studies.
??Metal surface cleaning is highly required in different fields of modern industry. Nuclear industry seeks for new methods for oxidized surface decontamination, and thermonuclear installations� (more)
Leontyev, Anton
Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report
This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed.
Hauptman, H.M.; Petro, J.N.; Jacobi, O. [and others
Reactor vessel support system. [LMFBR
Golden, M.P.; Holley, J.C.
DESIGN OF A MICROCHANNEL BASED SOLAR RECEIVER/REACTOR FOR METHANE-STEAM REFORMING Drost, K. J- current experimental data [1, 2]. Methane-steam reforming is modeled by three reduced-order reactions and modeling are used to investigate the strong endothermic reactions of methane-steam reforming inside
Apte, Sourabh V.
INL Experimental Program Roadmap for Thermal Hydraulic Code Validation
Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a �benchmark� database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related to VHTRs, sodium-cooled fast reactors, and light-water reactors. These experiments range from relatively low-cost benchtop experiments for investigating individual phenomena to large electrically-heated integral facilities for investigating reactor accidents and transients.
Glenn McCreery; Hugh McIlroy
Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements
The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.
J. D. Bess; T. L. Maddock; M. A. Marshall
Nuclear reactor construction with bottom supported reactor vessel
An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.
Sharbaugh, John E. (Bullskin Township, Fayette County, PA)
Design data for BORAX V are presented along with results of hazards evaluation studies. Considcration of the hazards associated with the operation of BORAX V was based on the following conditions: For normal steady-state power and experimental operation, the reactor and plant are adequately shielded and ventilated to allow personnel to be safely stationed in the turbine building and on the main floor of the reactor building. The control building is located one- half mile distant from the reactor building. For special, hazardous experiments, personnel are withdrawn from the reactor area. (M.C.G.)
Methane coupling by membrane reactor. Quarterly technical progress report: June 25, 1993--September 24, 1993
Several membranes have been investigated for use in a membrane reactor. Porous VYCOR has been tested for permeability changes with temperature. Three-sectional VYCOR membranes, with a porous central section, have been fabricated and tested in the experimental setup. Catalysts for methane coupling have been reviewed in the literature and five catalysts have been selected. Modeling studies of the methane oxidative coupling reaction in different reactor configurations shows higher C{sub 2} selectivity and yield with membrane reactors as compared to conventional packed bed reactors.
Ma, Yi Hua
Thermal-hydraulic aspects of flow inversion in a research reactor
PARET, a neutronics and thermal-hydraulics computer code, has been modified to account for natural convection in a reactor core. The code was then used to analyze the flow inversion that occurs in a reactor with heat removal by forced convection in the downward direction after a pump failure. Typical results are shown for a number of parameters. Research reactors normally operating much above ten MW are predicted to experience nucleate boiling in the event of a flow inversion. Comparison with experimental results from the Belgian BR2 reactor indicated general agreement although nucleate boiling that was analytically predicted was not noted in the BR2 data.
Smith, R.S.; Woodruff, W.L.
This homogeneous reactor comprises a core occupied by a solution of a fissile material in a moderator liquid and a breeder region enclosing the core and having a suspension of fertile material in the same moderator liquid. There is communication between the core and breeder to allow mass transfer and pressure equalization between the regions. The zones each have a separate circuit for removing heat by a mixer chamber situated inside the reactor vessel. The effluents coming from the two regions are mixed and led to a common device for separation into a clear solution and suspension, which are each led back to its corresponding circuit. To control the relative concentration of the two regions, an evaporator is provided separating a part of the moderator liquid from the solution occupying the core, the condensed separated moderator liquid being led into the breeder region. (NPO)
A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.
Jassby, D.L.
A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.
Fermi, E.; Szilard, L.
A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.
Hutter, E.
A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.
Peng, Yueng-Kay M. (Oak Ridge, TN)
Nuclear reactors of the homogeneous liquid fuel type are discussed. The reactor is comprised of an elongated closed vessel, vertically oriented, having a critical region at the bottom, a lower chimney structure extending from the critical region vertically upwardly and surrounded by heat exchanger coils, to a baffle region above which is located an upper chimney structure containing a catalyst functioning to recombine radiolyticallydissociated moderator gages. In operation the liquid fuel circulates solely by convection from the critical region upwardly through the lower chimney and then downwardly through the heat exchanger to return to the critical region. The gases formed by radiolytic- dissociation of the moderator are carried upwardly with the circulating liquid fuel and past the baffle into the region of the upper chimney where they are recombined by the catalyst and condensed, thence returning through the heat exchanger to the critical region.
Hammond, R.P.; Busey, H.M.
Water reactor fuel cladding
This patent describes a nuclear reactor fuel element cladding tube. It comprises: an outer cylindrical layer of a first zirconium alloy selected from the group consisting of Zircaloy-2 and Zircaloy-4; an inner cylindrical layer of a second zirconium alloy consisting essentially of about 0.19 to 0.6 wt.% tin, about 0.19 to less than 0.5 wt.% iron, about 100 to 700 ppm oxygen, less than 2000 ppm total impurities, and the remainder essentially zirconium; the inner layer characterized by aqueous corrosion resistance substantially the same as the first zirconium alloy; the inner layer characterized by improved resistance to PCI crack propagation under reactor operating conditions compared to the first zirconium alloy and substantially the same PCI crack propagation resistance compared to unalloyed zirconium; and the inner cylindrical layer is metallurgically bonded to the outer layer.
Foster, J.P.; McDonald, S.G.
Power Reactor Decommissioning Experience
During the past two decades the NRC regulated nuclear industry has encountered and dealt with a diverse range of political, financial and technological challenges while decommissioning its nuclear facilities. During that time, the decommissioning of nuclear facilities has evolved into a mature industry in the United States with a number of large power reactors successfully decommissioned and their NRC licenses terminated. One of the challenges discussed in this report is site release standards, required ...
A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)
Heckman, T.P.
Exhaust gas reactor
A reactor for the oxidation of unburned and partially burned components in the exhaust gas from an internal combustion engine comprising a chamber which is substantially circular in cross sections perpendicular to its axis, one or more inlet pipes which pass a mixture of exhaust gas and air substantially tangentially into the chamber near to one end thereof, and an outlet pipe near to the other end of the chamber and which is so arranged that exhaust gas leaves the chamber substantially tangentially. The tangential inlet and tangential outlet of gases minimizes energy losses in the gas passing through the reactor. The ratio of the cross-sectional areas of the inlet pipe(s) to reactor chamber is preferably from 1:9 to 25:36, and similar ranges of crosssectional area ratios are preferred for the outlet pipe and chamber. The ratio of the length of the reaction chamber to diameter is preferably from 1:1 to 4:1. The chamber may be cylindrical or divergent from inlet end to outlet end and may be thermally insulated.
Camarsa, M.; Cocchiara, F.; Garcea, G.P.
We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(\\alpha)$, in addition to solving a simple equilibrium problem in order to evaluate the numerical error behavior.
C. E. Michoski; J. A. Evans; P. G. Schmitz
Transforming criticality control methods for EBR-II fuel handling during reactor decommissioning
A review of the Department of Energy (DOE) request to decommission the Experimental Breeder Reactor-II (EBR-II) was conducted in order to develop a scope of work and analysis method for performing the safety review of the facility. Evaluation of the current national standards, DOE orders, EBR-II nuclear safeguards and criticality control practices showed that a decommissioning policy for maintaining criticality safety during a long term fuel transfer process did not exist. The purpose of this research was to provide a technical basis for transforming the reactor from an instrumentation and measurement controlled system to a system that provides both physical constraint and administrative controls to prevent criticality accidents. Essentially, this was done by modifying the reactor core configuration, reactor operations procedures and system instrumentation to meet the safety practices of ANS-8.1-1983. Subcritical limits were determined by applying established liquid metal reactor methods for both the experimental and computational validations.
Eberle, C.S.; Dean, E.M.; Angelo, P.L.
Solid oxide electrochemical reactor science.
Solid-oxide electrochemical cells are an exciting new technology. Development of solid-oxide cells (SOCs) has advanced considerable in recent years and continues to progress rapidly. This thesis studies several aspects of SOCs and contributes useful information to their continued development. This LDRD involved a collaboration between Sandia and the Colorado School of Mines (CSM) ins solid-oxide electrochemical reactors targeted at solid oxide electrolyzer cells (SOEC), which are the reverse of solid-oxide fuel cells (SOFC). SOECs complement Sandia's efforts in thermochemical production of alternative fuels. An SOEC technology would co-electrolyze carbon dioxide (CO{sub 2}) with steam at temperatures around 800 C to form synthesis gas (H{sub 2} and CO), which forms the building blocks for a petrochemical substitutes that can be used to power vehicles or in distributed energy platforms. The effort described here concentrates on research concerning catalytic chemistry, charge-transfer chemistry, and optimal cell-architecture. technical scope included computational modeling, materials development, and experimental evaluation. The project engaged the Colorado Fuel Cell Center at CSM through the support of a graduate student (Connor Moyer) at CSM and his advisors (Profs. Robert Kee and Neal Sullivan) in collaboration with Sandia.
Sullivan, Neal P. (Colorado School of Mines, Golden, CO); Stechel, Ellen Beth; Moyer, Connor J. (Colorado School of Mines, Golden, CO); Ambrosini, Andrea; Key, Robert J. (Colorado School of Mines, Golden, CO)
Applications for reactor-pumped lasers
Nuclear reactor-pumped lasers (RPLs) have been developed in the US by the Department of Energy for over two decades, with the primary research occurring at Sandia National Laboratories and Idaho National Engineering Laboratory. The US program has experimentally demonstrated reactor-pumped lasing in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1,271, 1,733, 1,792, 2,032, 2,630, 2,650, and 3,370 nm with intrinsic efficiency as high as 2.5%. The major strengths of a reactor-pumped laser are continuous high-power operation, modular construction, self-contained power, compact size, and a variety of wavelengths (from visible to infrared). These characteristics suggest numerous applications not easily accessible to other laser types. The continuous high power of an RPL opens many potential manufacturing applications such as deep-penetration welding and cutting of thick structures, wide-area hardening of metal surfaces by heat treatment or cladding application, wide-area vapor deposition of ceramics onto metal surfaces, production of sub-micron sized particles for manufacturing of ceramics, and 3-D ceramic lithography. In addition, a ground-based RPL could beam its power to space for such activities as illuminating geosynchronous communication satellites in the earth`s shadow to extend their lives, beaming power to orbital transfer vehicles, removing space debris, and providing power (from earth) to a lunar base during the long lunar night.
Lipinski, R.J.; McArthur, D.A. [Sandia National Labs., Albuquerque, NM (United States). Nuclear Systems Research
Design guide for category V reactors transient reactors
The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category V reactor structures, components, and systems.
Brynda, W J; Karol, R C; Lobner, P R; Powell, R W; Straker, E A
Computational Flow Predictions for the Lower Plenum of a High-Temperature, Gas-Cooled Reactor
Advanced gas-cooled reactors offer the potential advantage of higher efficiency and enhanced safety over present day nuclear reactors. Accurate simulation models of these Generation IV reactors are necessary for design and licensing. One design under consideration by the Very High Temperature Reactor (VHTR) program is a modular, prismatic gas-cooled reactor. In this reactor, the lower plenum region may experience locally high temperatures that can adversely impact the plant�s structural integrity. Since existing system analysis codes cannot capture the complex flow effects occurring in the lower plenum, computational fluid dynamics (CFD) codes are being employed to model these flows [1]. The goal of the present study is to validate the CFD calculations using experimental data.
Donna Post Guillen
Monitoring and Control Research Using a University Reactor and SBWR Test-Loop
The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs.
Robert M. Edwards
Convective heating analysis of an IFE target in a high temperature, low Reynolds number xenon environment
magnetic tokamak nuclear fusion reactor currently undersustain deuterium-tritium fusion reactors. Li + n ? T + He +for thermonuclear burn in fusion reactors is the Lawson
Holdener, Dain Steffen
Heavy Water and Graphite Reactors - Reactors designed/built by...
experiments, necessary to achieve higher precision for the determination of reactor power distribution patterns, effect of non-uniform void distributions, kinetic behavior,...
Reactor technology: power conversion systems and reactor operation and maintenance
The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He/sup 3/ reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored.
Powell, J.R.
Simulant-material experimental investigation of flow dynamics in the CRBR Upper-Core Structure
The results of a simulant-material experimental investigation of flow dynamics in the Clinch River Breeder Reactor (CRBR) Upper Core Structure are described. The methodology used to design the experimental apparatus and select test conditions is detailed. Numerous comparisons between experimental data and SIMMER-II Code calculations are presented with both advantages and limitations of the SIMMER modeling features identified.
Wilhelm, D.; Starkovich, V.S.; Chapyak, E.J.
Designing Reactors to Facilitate Decommissioning
Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.
Richard H. Meservey
Experimental Summary and Outlook
A brief experimental overview of the workshop is given, with emphasis on polarized targets from the experimental equipment perspective, and kinematic coverage, precision, and newly investigated channels from the experimental results perspective.
Peter Bosted
Indication of anomalous heat energy production in a reactor device
An experimental investigation of possible anomalous heat production in a special type of reactor tube named E-Cat HT is carried out. The reactor tube is charged with a small amount of hydrogen loaded nickel powder plus some additives. The reaction is primarily initiated by heat from resistor coils inside the reactor tube. Measurement of the produced heat was performed with high-resolution thermal imaging cameras, recording data every second from the hot reactor tube. The measurements of electrical power input were performed with a large bandwidth three-phase power analyzer. Data were collected in two experimental runs lasting 96 and 116 hours, respectively. An anomalous heat production was indicated in both experiments. The 116-hour experiment also included a calibration of the experimental set-up without the active charge present in the E-Cat HT. In this case, no extra heat was generated beyond the expected heat from the electric input. Computed volumetric and gravimetric energy densities were found to be far above those of any known chemical source. Even by the most conservative assumptions as to the errors in the measurements, the result is still one order of magnitude greater than conventional energy sources.
Giuseppe Levi; Evelyn Foschi; Torbj�rn Hartman; Bo H�istad; Roland Pettersson; Lars Tegn�r; Hanno Ess�n
Acceptability of reactors in space
Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it does not appear that reactors add measurably to the risk associated with the Space Transportation System.
Buden, D.
Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.
Dry cooling tower operating experience in the LOFT reactor
A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features.
Hunter, J.A.
MHD equilibrium properties of tokamak fusion reactor designs
The equilibrium properties of several Tokamak Reactor Designs are analyzed and compared for varying pressure and current profiles using the Princeton Equilibrium Code. It is found that the UWMAK configuration has a broader range of equilibria than the Princeton Reference Design configuration, but that the safety factor on axis is less than unity for peaked current distributions. The Argonne Experimental Power Reactor has a satisfactory range of equilibria, but a means of limiting or diverting the plasma has not yet been proposed, and this may substantially change the results obtained. (auth)
Todd, A. M.M.; Gralnick, S. L.; Dalhed, H. E.
Transport Reactor Facility
The Morgantown Energy Technology Center (METC) is currently evaluating hot gas desulfurization (HGD)in its on-site transport reactor facility (TRF). This facility was originally constructed in the early 1980s to explore advanced gasification processes with an entrained reactor, and has recently been modified to incorporate a transport riser reactor. The TRF supports Integrated Gasification Combined Cycle (IGCC) power systems, one of METC`s advanced power generation systems. The HGD subsystem is a key developmental item in reducing the cost and increasing the efficiency of the IGCC concept. The TRF is a unique facility with high-temperature, high-pressure, and multiple reactant gas composition capability. The TRF can be configured for reacting a single flow pass of gas and solids using a variety of gases. The gas input system allows six different gas inputs to be mixed and heated before entering the reaction zones. Current configurations allow the use of air, carbon dioxide, carbon monoxide, hydrogen, hydrogen sulfide, methane, nitrogen, oxygen, steam, or any mixture of these gases. Construction plans include the addition of a coal gas input line. This line will bring hot coal gas from the existing Fluidized-Bed Gasifier (FBG) via the Modular Gas Cleanup Rig (MGCR) after filtering out particulates with ceramic candle filters. Solids can be fed either by a rotary pocket feeder or a screw feeder. Particle sizes may range from 70 to 150 micrometers. Both feeders have a hopper that can hold enough solid for fairly lengthy tests at the higher feed rates, thus eliminating the need for lockhopper transfers during operation.
Berry, D.A.; Shoemaker, S.A.
A method and arrangement is presented for removing a fuel element from a neutronic reactor tube through which a liquid coolant is being circulaled. The fuel element is moved into a section of the tube beyond the reactor proper, and then the coolant in the tube between the fuel element and the reactor proper is frozen, so that the fuel element may be removed from the tube without loss of the coolant therein. The method is particularly useful in the case of a liquid metal- cooled reactor.
Reactor & Nuclear Systems Publications | ORNL
Publications and Reports NSED Monthly Reports Reactor and Nuclear Systems Publications 2013 Publications 2012 Publications 2011 Publications 2010 and Older Publications Nuclear...
Mirror Advanced Reactor Study (MARS)
Progress in a two year study of a 1200 MWe commercial tandem mirror reactor (MARS - Mirror Advanced Reactor Study) has reached the point where major reactor system technologies are identified. New design features of the magnets, blankets, plug heating systems and direct converter are described. With the innovation of radial drift pumping to maintain low plug density, reactor recirculating power fraction is reduced to 20%. Dominance of radial ion and impurity losses into the halo permits gridless, circular direct converters to be dramatically reduced in size. Comparisons of MARS with the Starfire tokamak design are made.
Logan, B.G.
When Do Commercial Reactors Permanently Shut Down?
Reports and Publications (EIA)
For those wishing to obtain current data, the following resources are available: U.S. reactors, go to EIA's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.
Information Center
Reactor operations informal report, October 1994
This monthly progress report is divided into two parts. Part one covers the Brookhaven Medical Research Reactor and part two covers the Brookhaven High Flux Beam Reactor. Information is given for each reactor covering the following areas: reactor operation; instrumentation; mechanical maintenance; occurrence reports; and reactor safety.
Hauptman, H.M.; Petro, J.N.; Jacobi, O.; Lettieri, V.; Holden, N.; Ports, D.; Petricek, R.
An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.
Leyse, C.F.; Putnam, G.E.
TRIGA reactor operating experience
The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections.
Anderson, T.V. [Oregon State University, Corvallis, OR (United States)
Reactor refueling containment system
A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.
Gillett, J.E.; Meuschke, R.E.
This report describes a method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.
Biparticle fluidized bed reactor
A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.
Scott, C.D.
A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)
Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.
This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.
Snell, A.H.
Biparticle fluidized bed