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Energy Return on Investment from Recycling Nuclear Fuel

Description: This report presents an evaluation of the Energy Return on Investment (EROI) from recycling an initial batch of 800 t/y of used nuclear fuel (UNF) through a Recycle Center under a number of different fuel cycle scenarios. The study assumed that apart from the original 800 t of UNF only depleted uranium was available as a feed. Therefore for each subsequent scenario only fuel that was derived from the previous fuel cycle scenario was considered. The scenarios represent a good cross section of the options available and the results contained in this paper and associated appendices will allow for other fuel cycle options to be considered.
Date: August 17, 2011
Partner: UNT Libraries Government Documents Department

Gamma-Free Neutron Detector Based upon Lithium Phosphate Nanoparticles

Description: A gamma-free neutron-sensitive scintillator is needed to enhance radiaition sensing and detection for nonproliferation applications. Such a scintillator would allow very large detectors to be placed at the perimeter of spent-fuel storage facilities at commercial nuclear power plants, so that any movement of spontaneously emitted neutrons from spent nuclear fuel or weapons grade plutonium would be noted in real-time. This task is to demonstrate that the technology for manufacturing large panels of fluor-doped plastic containing lithium-6 phosphate nanoparticles can be achieved. In order to detect neutrons, the nanoparticles must be sufficiently small so that the plastic remains transparent. In this way, the triton and alpha particles generated by the capture of the neutron will result in a photon burst that can be coupled to a wavelength shifting fiber (WLS) producing an optical signal of about ten nanoseconds duration signaling the presence of a neutron emitting source.
Date: August 28, 2007
Creator: Wallace, Steven
Partner: UNT Libraries Government Documents Department

Current Status of Deep Geological Repository Development

Description: This talk provided an overview of the current status of deep-geological-repository development worldwide. Its principal observation is that a broad consensus exists internationally that deep-geological disposal is the only long-term solution for disposition of highly radioactive nuclear waste. Also, it is now clear that the institutional and political aspects are as important as the technical aspects in achieving overall progress. Different nations have taken different approaches to overall management of their highly radioactive wastes. Some have begun active programs to develop a deep repository for permanent disposal: the most active such programs are in the United States, Sweden, and Finland. Other countries (including France and Russia) are still deciding on whether to proceed quickly to develop such a repository, while still others (including the UK, China, Japan) have affirmatively decided to delay repository development for a long time, typically for a generation of two. In recent years, a major conclusion has been reached around the world that there is very high confidence that deep repositories can be built, operated, and closed safely and can meet whatever safety requirements are imposed by the regulatory agencies. This confidence, which has emerged in the last few years, is based on extensive work around the world in understanding how repositories behave, including both the engineering aspects and the natural-setting aspects, and how they interact together. The construction of repositories is now understood to be technically feasible, and no major barriers have been identified that would stand in the way of a successful project. Another major conclusion around the world is that the overall cost of a deep repository is not as high as some had predicted or feared. While the actual cost will not be known in detail until the costs are incurred, the general consensus is that the total life-cycle cost will not ...
Date: August 29, 2005
Creator: Budnitz, R. J.
Partner: UNT Libraries Government Documents Department

ATWS Transients for the 2400 MWt Gas-Cooled Fast Reactor

Description: Reactivity transients have been analyzed with an updated RELAPS-3D (ver. 2.4.2) system model of the pin core design for the 2400MWt gas-cooled fast reactor (GCFR). Additional reactivity parameters were incorporated in the RELAP5 point-kinetics model to account for reactivity feedbacks due to axial and radial expansion of the core, fuel temperature changes (Doppler effect), and pressure changes (helium density changes). Three reactivity transients without scram were analyzed and the incidents were initiated respectively by reactivity ramp, loss of load, and depressurization. During the course of the analysis the turbine bypass model for the power conversion unit (PCU) was revised to enable a better utilization of forced flow cooling after the PCU is tripped. The analysis of the reactivity transients demonstrates the significant impact of the PCU on system pressure and core flow. Results from the modified turbine bypass model suggest a success path for the GCFR to mitigate reactivity transients without scram.
Date: August 5, 2007
Creator: Cheng,L.Y. & Ludewig, H.
Partner: UNT Libraries Government Documents Department

Environmentally assisted cracking in light water reactors annual report January - December 2005.

Description: This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth ...
Date: August 31, 2007
Creator: Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J. et al.
Partner: UNT Libraries Government Documents Department

Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

Description: The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life ({var_epsilon}-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of fatigue crack initiation in austenitic stainless steels in LWR coolant environments. The existing fatigue {var_epsilon}-N data have been evaluated to establish the effects of key material, loading, and environmental parameters (such as steel type, strain range, strain rate, temperature, dissolved-oxygen level in water, and flow rate) on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. The influence of reactor environments on the mechanism of fatigue crack initiation in these steels is also discussed.
Date: August 1, 2002
Creator: Chopra, O. K. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Deep Burn Develpment of Transuranic Fuel for High-Temperature Helium-Cooled Reactors - July 2010

Description: The DB Program Quarterly Progress Report for April - June 2010, ORNL/TM/2010/140, was distributed to program participants on August 4. This report discusses the following: (1) TRU (transuranic elements) HTR (high temperature helium-cooled reactor) Fuel Modeling - (a) Thermochemical Modeling, (b) 5.3 Radiation Damage and Properties; (2) TRU HTR Fuel Qualification - (a) TRU Kernel Development, (b) Coating Development, (c) ZrC Properties and Handbook; and (3) HTR Fuel Recycle - (a) Recycle Processes, (b) Graphite Recycle, (c) Pyrochemical Reprocessing - METROX (metal recovery from oxide fuel) Process Development.
Date: August 1, 2010
Creator: Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D & Bell, Gary L
Partner: UNT Libraries Government Documents Department

NRC Job Code V6060: Extended in-situ and real time monitoring. Task 4: Detection and monitoring of leaks at nuclear power plants external to structures

Description: In support of Task 4 of the NRC study on compliance with 10 CFR part 20.1406, minimization of contamination, Argonne National Laboratory (ANL) conducted a one-year scoping study, in concert with a parallel study performed by NRC/NRR staff, on monitoring for leaks at nuclear power plants (NPPs) external to structures. The objective of this task-4 study is to identify and assess those sensors and monitoring techniques for early detection of abnormal radioactive releases from the engineered facility structures, systems and components (SSCs) to the surrounding underground environment in existing NPPs and planned new reactors. As such, methods of interest include: (1) detection of anomalous water content of soils surrounding SSCs, (2) radionuclides contained in the leaking water, and (3) secondary signals such as temperature. ANL work scope includes mainly to (1) identify, in concert with the nuclear industry, the sensors and techniques that have most promise to detect radionuclides and/or associated chemical releases from SSCs of existing NPPs and (2) review and provide comments on the results of the NRC/NRR staff scoping study to identify candidate technologies. This report constitutes the ANL deliverable of the task-4 study. It covers a survey of sensor technologies and leak detection methods currently applied to leak monitoring at NPPs. The survey also provides a technology evaluation that identifies their strength and deficiency based on their detection speed, sensitivity, range and reliability. Emerging advanced technologies that are potentially capable of locating releases, identifying the radionuclides, and estimating their concentrations and distributions are also included in the report along with suggestions of required further research and development.
Date: August 1, 2012
Creator: Sheen, S. H. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Lifecycle Prognostics Architecture for Selected High-Cost Active Components

Description: There are an extensive body of knowledge and some commercial products available for calculating prognostics, remaining useful life, and damage index parameters. The application of these technologies within the nuclear power community is still in its infancy. Online monitoring and condition-based maintenance is seeing increasing acceptance and deployment, and these activities provide the technological bases for expanding to add predictive/prognostics capabilities. In looking to deploy prognostics there are three key aspects of systems that are presented and discussed: (1) component/system/structure selection, (2) prognostic algorithms, and (3) prognostics architectures. Criteria are presented for component selection: feasibility, failure probability, consequences of failure, and benefits of the prognostics and health management (PHM) system. The basis and methods commonly used for prognostics algorithms are reviewed and summarized. Criteria for evaluating PHM architectures are presented: open, modular architecture; platform independence; graphical user interface for system development and/or results viewing; web enabled tools; scalability; and standards compatibility. Thirteen software products were identified and discussed in the context of being potentially useful for deployment in a PHM program applied to systems in a nuclear power plant (NPP). These products were evaluated by using information available from company websites, product brochures, fact sheets, scholarly publications, and direct communication with vendors. The thirteen products were classified into four groups of software: (1) research tools, (2) PHM system development tools, (3) deployable architectures, and (4) peripheral tools. Eight software tools fell into the deployable architectures category. Of those eight, only two employ all six modules of a full PHM system. Five systems did not offer prognostic estimates, and one system employed the full health monitoring suite but lacked operations and maintenance support. Each product is briefly described in Appendix A. Selection of the most appropriate software package for a particular application will depend on the chosen component, system, or structure. ...
Date: August 1, 2011
Creator: Lybeck, N.; Pham, B.; Tawfik, M.; Coble, J. B.; Meyer, R. M.; Ramuhalli, P. et al.
Partner: UNT Libraries Government Documents Department

Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community

Description: Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.
Date: August 1, 2011
Creator: Lybeck, Nancy J.; Tawfik, Magdy S. & Pham, Binh T.
Partner: UNT Libraries Government Documents Department

AGR-2 Data Qualification Report for ATR Cycles 147A, 148A, 148B, and 149A

Description: This report presents the data qualification status of fuel irradiation data from the first four reactor cycles (147A, 148A, 148B, and 149A) of the on-going second Advanced Gas Reactor (AGR-2) experiment as recorded in the NGNP Data Management and Analysis System (NDMAS). This includes data received by NDMAS from the period June 22, 2010 through May 21, 2011. AGR-2 is the second in a series of eight planned irradiation experiments for the AGR Fuel Development and Qualification Program, which supports development of the very high temperature gas-cooled reactor (VHTR) under the Next Generation Nuclear Plant (NGNP) Project. Irradiation of the AGR-2 test train is being performed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and is planned for 600 effective full power days (approximately 2.75 calendar years) (PLN-3798). The experiment is intended to demonstrate the performance of UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Data qualification status of the AGR-1 experiment was reported in INL/EXT-10-17943 (Abbott et al. 2010).
Date: August 1, 2011
Creator: Abbott, Michael L. & Daum, Keith A.
Partner: UNT Libraries Government Documents Department

High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

Description: This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.
Date: August 1, 2011
Creator: Demick, Larry
Partner: UNT Libraries Government Documents Department

Modular HTGR Safety Basis and Approach

Description: The Next Generation Nuclear Plant (NGNP) will be a licensed commercial high temperature gas-cooled reactor (HTGR) capable of producing electricity and/or high temperature process heat for industrial markets supporting a range of end-user applications. The NGNP Project has adopted the 10 CFR 52 Combined License (COL) process, as recommended in the NGNP Licensing Strategy - A Report to Congress, dated August 2008, as the foundation for the NGNP licensing strategy [DOE/NRC 2008]. Nuclear Regulatory Commission (NRC) licensing of the NGNP plant utilizing this process will demonstrate the efficacy for licensing future HTGRs for commercial industrial applications. This information paper is one in a series of submittals that address key generic issues of the priority licensing topics as part of the process for establishing HTGR regulatory requirements. This information paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach with the NRC staff and public stakeholders. The NGNP project does not expect to receive comments on this information paper because other white papers are addressing key generic issues of the priority licensing topics in greater detail.
Date: August 1, 2011
Creator: Hicks, Thomas
Partner: UNT Libraries Government Documents Department

Conceptual Design of a CERMET NTR Fission Core Using Multiphysics Modeling Techniques

Description: An initial pre-conceptual CERMET Nuclear Thermal Propulsion reactor system is investigated within this paper. Reactor configurations are investigated where the fuel consists of 60 vol.% UO2 and 40 vol.% W where the UO2 consists of Gd2O3 concentrations of 5 and 10 mol.%.Gd2O3. The fuel configuration consisting of 5 mol.% UO2 was found to have a total mass of 2761 kg and a thrust to weight ratio of 4.10 and required a coolant channel surface area to fueled volume ratio of approximately 15.0 in order to keep the centerline temperature below 3000 K. The configuration consisting of 10 mol.% Gd2O3 required a surface area to volume ratio of approximately 12.2 to cool the reactor to a peak temperature of 3000 K and had a total mass of 3200 kg and a thrust to weight ratio of 3.54. It is not known yet what concentration of Gd2O3 is required to maintain fuel stability at 3000 K; however, both reactors offer the potential for operations at 25,000 lb, and at a specific impulse which may range from 900 to 950 seconds.
Date: August 1, 2011
Creator: Webb, Jonathan A.; Gross, Brian J. & Taitano, William T.
Partner: UNT Libraries Government Documents Department

OM Code Requirements For MOVs -- OMN-1 and Appendix III

Description: The purpose or scope of the ASME OM Code is to establish the requirements for pre-service and in-service testing of nuclear power plant components to assess their operational readiness. For MOVs this includes those that perform a specific function in shutting down a reactor to the safe shutdown condition, maintaining the safe shutdown condition, and mitigating the consequences of an accident. This paper will present a brief history of industry and regulatory activities related to MOVs and the development of Code requirements to address weaknesses in earlier versions of the OM Code. The paper will discuss the MOV requirements contained in the 2009 version of ASME OM Code, specifically Mandatory Appendix III and OMN-1, Revision 1.
Date: August 1, 2011
Creator: DeWall, Kevin G.
Partner: UNT Libraries Government Documents Department

Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

Description: This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.
Date: August 1, 2012
Creator: Ilas, Germina; Ilas, Dan; Kelly, Ryan P & Sunny, Eva E
Partner: UNT Libraries Government Documents Department

Neutron Spectra and Dose Equivalent Inside Nuclear Power Reactor Containment

Description: This study was conducted to determine absorbed dose, dose-equivalent rates, and neutron spectra inside containment at nuclear power plants. We gratefully acknowledge funding support by the Nuclear Regulatory Commission. The purpose of this study is: 1) measure dose-equivalent rates with various commercial types of rem meters, such as the Snoopy and Rascal, and neutron absorbed dose rates with a tissue-equivalent proportional counter 2) determine neutron spectra using the multi sphere or Bonner sphere technique and a helium-3 spectrometer 3) compare several types of personnel neutron dosimeter responses such as NTA film, polycarbonates, TLD albedo, and a recently introduced proton recoil track etch dosimeter, and CR-39. These measurements were made inside containments of pressurized water reactors (PWRs) and outside containment penetrations of boiling water reactors (BWRs) operating at full power. The neutron spectral information, absorbed dose. and dose-equivalent measurements are needed for proper interpretation of instrument and personnel dosimeter responses.
Date: August 1, 1981
Creator: Aldrich, J. M.
Partner: UNT Libraries Government Documents Department

An Investigation of the Application of the Gas Generator-Free Turbine Cycle to a Nuclear Powered Aircraft

Description: This study has investigated the feasibility of installing a gas generator-free turbine type power plant in the R3Y aircraft, using a circulating fuel reactor as a power source. Two variations of the cycle were considered. The split flow cycle bleeds high temperature, high pressure air from the gas generator directly to the free turbine in the wing. The through flow cycle partially expands the high temperature, high pressure air through the compressor turbine of the gas generator then directs the compressor-turbine exhaust air to the free turbine in the wing. Design parameters of pressure ratio, radiator depth, radiation flow density, and hot gas duct size were optimized to give minimum weight per shaft horsepower of the complete power plant. The weight of a split flow power plant capable of supplying 22,000 shaft horsepower was found to be 116,600 pounds. The weight of a similar through flow power plant was found to be 119,900 pounds. The reactor power required in both cases was 70 megawatts. The nominal gross weight of the R3Y airplane is 175,000 pounds. With pay loads of approximately 20,000 pounds, either nuclear conversion will have a gross weight of 200,000 pounds. It was found that either cycle could be installed in the R3Y aircraft; however, the installation of either would require major structural redesign. The split flow cycle with its smaller hot air ducts required the least amount of redesign. A comparison of existing aircraft engines with a preliminary design of the split flow turbo-components indicated that the compressor and possibly the free turbine could be adapted from current engine components.
Date: August 1, 1957
Creator: Alvis, J. H. & Chessman, S. R.
Partner: UNT Libraries Government Documents Department

Westinghouse Reactor Protection System Unavailability, 1984--1995

Description: An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U. S. Westinghouse commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.
Date: August 1, 1999
Creator: Eide, Steven Arvid; Calley, Michael Brennan; Gentillon, Cynthia Ann; Wierman, Thomas Edward; Rasmuson, D. & Marksberry, D.
Partner: UNT Libraries Government Documents Department

General Electric Reactor Protection System Unavailability, 1984--1995

Description: An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U. S. General Electric commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.
Date: August 1, 1999
Creator: Eide, Steven Arvid; Calley, Michael Brennan; Gentillon, Cynthia Ann; Wierman, Thomas Edward; Hamzehee, H. & Rasmuson, D.
Partner: UNT Libraries Government Documents Department

NARAC: An Emergency Response Resource for Predicting the Atmospheric Dispersion and Assessing the Consequences of Airborne Radionuclides

Description: Hazardous radioactive materials can be released into the atmosphere by accidents at nuclear power plants, fuel processing facilities, and other facilities, and by transportation accidents involving nuclear materials. In addition, the post-cold-war proliferation of nuclear material has increased the potential for terrorism scenarios involving radiological dispersal devices, improvised nuclear devices, and inadequately secured military nuclear weapons. To mitigate these risks, the National Atmospheric Release Advisory Center (NARAC) serves as a national resource for the United States, providing tools and services to quickly predict the environmental contamination and health effects caused by airborne radionuclides, and to provide scientifically based guidance to emergency managers for the protection of human life. NARAC's expert staff uses computer models, supporting databases, software systems, and communications systems to predict the plume paths and consequences of radiological, chemical, and biological atmospheric releases.
Date: August 23, 2005
Creator: Bradley, M M
Partner: UNT Libraries Government Documents Department

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

Description: The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.
Date: August 1, 2009
Creator: Rosenthal, Murray Wilford
Partner: UNT Libraries Government Documents Department