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Thorium Energy Futures

Description: The potential for thorium as an alternative or supplement to uranium in fission power generation has long been recognised, and several reactors, of various types, have already operated using thorium-based fuels. Accelerator Driven Subcritical (ADS) systems have benefits and drawbacks when compared to conventional critical thorium reactors, for both solid and molten salt fuels. None of the four options - liquid or solid, with or without an accelerator - can yet be rated as better or worse than the other three, given today's knowledge. We outline the research that will be necessary to lead to an informed choice.
Date: July 2012
Creator: Peggs, Stephen; Roser, Thomas; Parks, G; Lindroos, Mats; Seviour, Rebecca; Henderson, Stuart et al.
Partner: UNT Libraries Government Documents Department

DEVELOPMENT OF HFE SECTIONS OF DG-1145.

Description: For the licensing of the current fleet of commercial nuclear power plants (NPPs), the Nuclear Regulatory Commission (NRC) used two key documents, NUREG-0800 and Regulatory Guide (RG) 1.70. RG 1.70 provided guidance to applicants on the contents needed in their Safety Analysis Reports (SARs) submitted as part of their application to construct or operate an NPP. NUREG-0800, the NRC Standard Review Plan (SRP), provides guidance to the NRR staff reviewers on performing their safety reviews of these applications. As part of the preparation for a new wave of improved NPP designs the NRC is in the process of updating the SRP and is also developing a new RG designated as draft RG or DG-1145, ''Combined License Applications for Nuclear Power Plants (LWR Edition).'' This will eventually become RG 1.206 and will take the place of RG 1.70. This will provide guidance for combined license (COL) applicants, as well as for other 10CFR Part 52 variations that are permitted.
Date: March 26, 2007
Creator: HIGGINS,J.C.; OHARA, J.M. & BONGARRA, J.
Partner: UNT Libraries Government Documents Department

FINITE ELEMENT ANALYSIS OF JNES/NUPEC SEISMIC SHEAR WALL CYCLIC AND SHAKING TABLE TEST DATA.

Description: This paper describes a finite element analysis to predict the JNES/NUPEC cyclic and shaking table RC shear wall test data, as part of a collaborative agreement between the U.S. NRC and JNES to study seismic issues important to the safe operation of commercial nuclear power plant (NPP) structures, systems and components (SSC). The analyses described in this paper were performed using ANACAP reinforced concrete models. The paper describes the ANACAP analysis models and discusses the analysis comparisons with the test data. The ANACAP capability for modeling nonlinear cyclic characteristics of reinforced concrete shear wall structures was confirmed by the close comparisons between the ANACAP analysis results and the JNES/NUPEC cyclic test data. Reasonable agreement between the analysis results and the test data was demonstrated for the hysteresis loops and the shear force orbits, in terms of both the overall shape and the cycle-to-cycle comparisons. The ANACAP simulation analysis of the JNES/NUPEC shaking table test was also performed, which demonstrated that the ANACAP dynamic analysis with concrete material model is able to capture the progressive degrading behavior of the shear wall as indicated from the test data. The ANACAP analysis also predicted the incipient failure of the shear wall, reasonably close to the actual failure declared for the test specimen. In summary, the analyses of the JNES/NUPEC cyclic and shaking table RC shear wall tests presented in this paper have demonstrated the state-of-the-art analysis capability for determining the seismic capacity of RC shear wall structures.
Date: April 12, 2007
Creator: XU,J.; NIE, J.; HOFMAYER, C. & ALI, S.
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

Flooding Experiments and Modeling for Improved Reactor Safety

Description: Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.
Date: September 14, 2008
Creator: Solmos, M., Hogan, K.J., VIerow, K.
Partner: UNT Libraries Government Documents Department

Nonlinear Seismic Correlation Analysis of the JNES/NUPEC Large-Scale Piping System Tests.

Description: The Japan Nuclear Energy Safety Organization/Nuclear Power Engineering Corporation (JNES/NUPEC) large-scale piping test program has provided valuable new test data on high level seismic elasto-plastic behavior and failure modes for typical nuclear power plant piping systems. The component and piping system tests demonstrated the strain ratcheting behavior that is expected to occur when a pressurized pipe is subjected to cyclic seismic loading. Under a collaboration agreement between the US and Japan on seismic issues, the US Nuclear Regulatory Commission (NRC)/Brookhaven National Laboratory (BNL) performed a correlation analysis of the large-scale piping system tests using derailed state-of-the-art nonlinear finite element models. Techniques are introduced to develop material models that can closely match the test data. The shaking table motions are examined. The analytical results are assessed in terms of the overall system responses and the strain ratcheting behavior at an elbow. The paper concludes with the insights about the accuracy of the analytical methods for use in performance assessments of highly nonlinear piping systems under large seismic motions.
Date: June 1, 2008
Creator: Nie,J.; DeGrassi, G.; Hofmayer, C. & Ali, S.
Partner: UNT Libraries Government Documents Department

MARKOV Model Application to Proliferation Risk Reduction of an Advanced Nuclear System

Description: The Generation IV International Forum (GIF) emphasizes proliferation resistance and physical protection (PR&PP) as a main goal for future nuclear energy systems. The GIF PR&PP Working Group has developed a methodology for the evaluation of these systems. As an application of the methodology, Markov model has been developed for the evaluation of proliferation resistance and is demonstrated for a hypothetical Example Sodium Fast Reactor (ESFR) system. This paper presents the case of diversion by the facility owner/operator to obtain material that could be used in a nuclear weapon. The Markov model is applied to evaluate material diversion strategies. The following features of the Markov model are presented here: (1) An effective detection rate has been introduced to account for the implementation of multiple safeguards approaches at a given strategic point; (2) Technical failure to divert material is modeled as intrinsic barriers related to the design of the facility or the properties of the material in the facility; and (3) Concealment to defeat or degrade the performance of safeguards is recognized in the Markov model. Three proliferation risk measures are calculated directly by the Markov model: the detection probability, technical failure probability, and proliferation time. The material type is indicated by an index that is based on the quality of material diverted. Sensitivity cases have been done to demonstrate the effects of different modeling features on the measures of proliferation resistance.
Date: July 13, 2008
Creator: Bari,R.A.
Partner: UNT Libraries Government Documents Department

MONITORING WASTE HEAT REJECTION TO THE ENVIRONMENT VIA REMOTE SENSING

Description: Nuclear power plants typically use waste heat rejection systems such as cooling lakes and natural draft cooling towers. These systems are designed to reduce cooling water temperatures sufficiently to allow full power operation even during adverse meteorological conditions. After the power plant is operational, the performance of the cooling system is assessed. These assessments usually rely on measured temperatures of the cooling water after it has lost heat to the environment and is being pumped back into the power plant (cooling water inlet temperature). If the cooling system performance is not perceived to be optimal, the utility will collect additional data to determine why. This paper discusses the use of thermal imagery collected from aircraft and satellites combined with numerical simulation to better understand the dynamics and thermodynamics of nuclear power plant waste heat dissipation systems. The ANS meeting presentation will discuss analyses of several power plant cooling systems based on a combination of remote sensing data and hydrodynamic modeling.
Date: January 13, 2009
Creator: Garrett, A
Partner: UNT Libraries Government Documents Department

HIGHLY ENRICHED URANIUM BLEND DOWN PROGRAM AT THE SAVANNAH RIVER SITE PRESENT AND FUTURE

Description: The Department of Energy (DOE) and Tennessee Valley Authority (TVA) entered into an Interagency Agreement to transfer approximately 40 metric tons of highly enriched uranium (HEU) to TVA for conversion to fuel for the Browns Ferry Nuclear Power Plant. Savannah River Site (SRS) inventories included a significant amount of this material, which resulted from processing spent fuel and surplus materials. The HEU is blended with natural uranium (NU) to low enriched uranium (LEU) with a 4.95% 235U isotopic content and shipped as solution to the TVA vendor. The HEU Blend Down Project provided the upgrades needed to achieve the product throughput and purity required and provided loading facilities. The first blending to low enriched uranium (LEU) took place in March 2003 with the initial shipment to the TVA vendor in July 2003. The SRS Shipments have continued on a regular schedule without any major issues for the past 5 years and are due to complete in September 2008. The HEU Blend program is now looking to continue its success by dispositioning an additional approximately 21 MTU of HEU material as part of the SRS Enriched Uranium Disposition Project.
Date: June 5, 2008
Creator: Magoulas, V; Charles Goergen, C & Ronald Oprea, R
Partner: UNT Libraries Government Documents Department

EFFECT OF ELECTROLYZER CONFIGURATION AND PERFORMANCE ON HYBRID SULFUR PROCESS NET THERMAL EFFICIENCY

Description: Hybrid Sulfur cycle is gaining popularity as a possible means for massive production of hydrogen from nuclear energy. Several different ways of carrying out the SO{sub 2}-depolarized electrolysis step are being pursued by a number of researchers. These alternatives are evaluated with complete flowsheet simulations and on a common design basis using Aspen Plus{trademark}. Sensitivity analyses are performed to assess the performance potential of each configuration, and the flowsheets are optimized for energy recovery. Net thermal efficiencies are calculated for the best set of operating conditions for each flowsheet and the results compared. This will help focus attention on the most promising electrolysis alternatives. The sensitivity analyses should also help identify those features that offer the greatest potential for improvement.
Date: March 16, 2007
Creator: Gorensek, M
Partner: UNT Libraries Government Documents Department

AFCI-2.0 Library of Neutron Cross Section Covariances

Description: Neutron cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The primary purpose of the library is to provide covariances for the Advanced Fuel Cycle Initiative (AFCI) data adjustment project, which is focusing on the needs of fast advanced burner reactors. The covariances refer to central values given in the 2006 release of the U.S. neutron evaluated library ENDF/B-VII. The preliminary version (AFCI-2.0beta) has been completed in October 2010 and made available to the users for comments. In the final 2.0 release, covariances for a few materials were updated, in particular new LANL evaluations for {sup 238,240}Pu and {sup 241}Am were adopted. BNL was responsible for covariances for structural materials and fission products, management of the library and coordination of the work, while LANL was in charge of covariances for light nuclei and for actinides.
Date: June 26, 2011
Creator: Herman, M.; Herman,M.; Oblozinsky,P.; Mattoon,C.; Pigni,M.; Hoblit,S. et al.
Partner: UNT Libraries Government Documents Department

HANFORDS PUBLIC TOUR PROGRAM - AN EXCELLENT EDUCATIONAL TOOL

Description: Prior to 2001, the Department of Energy (DOE) sponsored limited tours of the Hanford Site for the public, but discontinued the program after the 9/11 terrorist attacks on the U.S. In 2003, DOE's Richland Operations Office (DOE-RL) requested the site's prime contractor to reinstate the public tour program starting in 2004 under strict controls and security requirements. The planning involved a collaborative effort among the security, safety and communications departments of DOE-RL and the site's contracting companies. This paper describes the evolution of, and enhancements to, Hanford's public tours, including the addition of a separate tour program for the B Reactor, the first full-scale nuclear reactor in the world. Topics included in the discussion include the history and growth of the tour program, associated costs, and visitor surveys and assessments.
Date: December 7, 2010
Creator: KM, SINCLAIR
Partner: UNT Libraries Government Documents Department

CLOSURE OF THE FAST FLUX TEST FACILITY (FFTF) CURRENT STATUS & FUTURE PLANS

Description: The Fast Flux Test Facility (FFTF) was a 400 MWt sodium-cooled fast reactor situated on the U.S. Department of Energy's (DOE) Hanford Site in the southeastern portion of Washington State. DOE issued the final order to shut down the facility in 2001, when it was concluded that there was no longer a need for FFTF. Deactivation activities are in progress to remove or stabilize major hazards and deactivate systems to achieve end points documented in the project baseline. The reactor has been defueled, and approximately 97% of the fuel has been removed from the facility. Approximately 97% of the sodium has been drained from the plant's systems and placed into an on-site Sodium Storage Facility. The residual sodium will be kept frozen under a blanket of inert gas until it is removed later as part of the facility's decontamination and decommissioning (D&D). Plant systems have been shut down and placed in a low-risk state to minimize requirements for surveillance and maintenance. D&D work cannot begin until an Environmental Impact Statement has been prepared to evaluate various end state options and to provide a basis for selecting one of the options. The Environmental Impact Statement is expected to be issued in 2009.
Date: May 23, 2007
Creator: LESPERANCE, C.P.
Partner: UNT Libraries Government Documents Department

NUCLEAR DATA NEEDS FOR ADVANCED REACTOR SYSTEMS. A NEA NUCLEAR SCIENCE COMMITTEE INITIATIVE.

Description: The Working Party on Evaluation Cooperation (WPEC) of the OECD Nuclear Energy Agency Nuclear Science Committee has established an International Subgroup to perform an activity in order to develop a systematic approach to define data needs for Gen-IV and, in general, for advanced reactor systems. A methodology, based on sensitivity analysis has been agreed and representative core configurations for Sodium, Gas and Lead cooled Fast Reactors (SFR, GFR, LFR) have been defined as well as a high burn-up VHTR and a high burn-up PWR. In the case of SFRs, both a TRU burner (called in fact SFR) and a core configuration with homogeneous recycling of not separated TRU (called EFR) have been considered.
Date: April 22, 2007
Creator: SALVATORES,J.M.; ALIBERTI, G.; PALMIOTTI, G.; ROCHMAN, D.; OBLOZINSKY, P.; HERMANN, M. et al.
Partner: UNT Libraries Government Documents Department

NEW ENDF/B-VII.0 EVALUATIONS OF NEUTRON CROSS SECTIONS FOR 32 FISSION PRODUCTS.

Description: Neutron cross sections for fission products play important role not only in the design of extended burnup core and fast reactors, but also in the study of the backend fuel cycle and the criticality analysis of spent fuel. New evaluations in both the resonance and fast neutron regions were performed by the KAERI-BNL collaboration for 32 fission products. These were {sup 95}Mo, {sup 101}Ru, {sup 103}Rh, {sup 105}Pd, {sup 109}Ag, {sup 131}Xe, {sup 133}Cs, {sup 141}Pr, and complete isotope chains of {sup 142-148,150}Nd, {sup 144,147,148-154}Sm, and {sup 156,158,160-164}Dy. The evaluations cover a large amount of reaction channels, including all those needed for neutronics calculations. Also, they cover the entire energy range, from 10{sup -5} eV to 20 MeV, including the thermal, resolved, and unresolved resonance regions, and the fast neutron region.
Date: April 22, 2007
Creator: KIM,H.; LEE, Y.-O.; HERMAN, M.; MUGHABGHAB, S.F.; OBLOZINSKY, P. & ROCHMAN, D.
Partner: UNT Libraries Government Documents Department

OVERVIEW ON BNL ASSESSMENT OF SEISMIC ANALYSIS METHODS FOR DEEPLY EMBEDDED NPP STRUCTURES.

Description: A study was performed by Brookhaven National Laboratory (BNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (USNRC), to determine the applicability of established soil-structure interaction analysis methods and computer programs to deeply embedded and/or buried (DEB) nuclear power plant (NPP) structures. This paper provides an overview of the BNL study including a description and discussions of analyses performed to assess relative performance of various SSI analysis methods typically applied to NPP structures, as well as the importance of interface modeling for DEB structures. There are four main elements contained in the BNL study: (1) Review and evaluation of existing seismic design practice, (2) Assessment of simplified vs. detailed methods for SSI in-structure response spectrum analysis of DEB structures, (3) Assessment of methods for computing seismic induced earth pressures on DEB structures, and (4) Development of the criteria for benchmark problems which could be used for validating computer programs for computing seismic responses of DEB NPP structures. The BNL study concluded that the equivalent linear SSI methods, including both simplified and detailed approaches, can be extended to DEB structures and produce acceptable SSI response calculations, provided that the SSI response induced by the ground motion is very much within the linear regime or the non-linear effect is not anticipated to control the SSI response parameters. The BNL study also revealed that the response calculation is sensitive to the modeling assumptions made for the soil/structure interface and application of a particular material model for the soil.
Date: April 1, 2007
Creator: XU,J.; COSTANTINO, C.; HOFMAYER, C. & GRAVES, H.
Partner: UNT Libraries Government Documents Department

Further Dosimetry Studies at Rhode Island Nuclear Science Center.

Description: The RINSC is a 2 mega-watt, light water and graphite moderated and cooled reactor that has a graphite thermal column built as a user facility for sample irradiation. Over the past decade, after the reactor conversion from a highly-enriched uranium core to a low-enriched one, flux and dose measurements and calculations had been performed in the thermal column to update the ex-core parameters and to predict the effect from in-core fuel burn-up and rearrangement. The most recent data from measurements and calculations that have been made at the RINSC thermal column since October of 2005 are reported.
Date: May 5, 2008
Creator: Reciniello,R.N.; Holden, N.E.; Hu, J.-P.; Johnson, D.G.; Meddleton, M. & Tehan, T.N.
Partner: UNT Libraries Government Documents Department

THE DEACTIVATION DECONTAMINATION & DECOMMISSIONING OF THE PLUTONIUM FINISHING PLANT (PFP) A FORMER PLUTONIUM PROCESSING FACILITY AT DOE HANFORD SITE

Description: The Plutonium Finishing Plant (PFP) was constructed as part of the Manhattan Project during World War II. The Manhattan Project was developed to usher in the use of nuclear weapons to end the war. The primary mission of the PFP was to provide plutonium used as special nuclear material (SNM) for fabrication of nuclear devices for the war effort. Subsequent to the end of World War II, the PFP's mission expanded to support the Cold War effort through plutonium production during the nuclear arms race and later the processing of fuel grade mixed plutonium-uranium oxide to support DOE's breeder reactor program. In October 1990, at the close of the production mission for PFP, a shutdown order was prepared by the Department of Energy (DOE) in Washington, DC and issued to the Richland DOE field office. Subsequent to the shutdown order, a team from the Defense Nuclear Facilities Safety Board (DNFSB) analyzed the hazards at PFP associated with the continued storage of certain forms of plutonium solutions and solids. The assessment identified many discrete actions that were required to stabilize the different plutonium forms into stable form and repackage the material in high integrity containers. These actions were technically complicated and completed as part of the PFP nuclear material stabilization project between 1995 and early 2005. The completion of the stabilization project was a necessary first step in deactivating PFP. During stabilization, DOE entered into negotiations with the U.S. Environmental Protection Agency (EPA) and the State of Washington and established milestones for the Deactivation and Decommissioning (D&D) of the PFP. The DOE and its contractor, Fluor Hanford (Fluor), have made great progress in deactivating, decontaminating and decommissioning the PFP at the Hanford Site as detailed in this paper. Background information covering the PFP D&D effort includes descriptions of negotiations with the ...
Date: February 1, 2006
Creator: CHARBONEAU, S.L.
Partner: UNT Libraries Government Documents Department

Update on US High Density Fuel Fabrication Development

Description: Second generation uranium molybdenum fuel has shown excellent in-reactor irradiation performance. This metallic fuel type is capable of being fabricated at much higher loadings than any presently used research reactor fuel. Due to the broad range of fuel types this alloy system encompasses—fuel powder to monolithic foil and binary fuel systems to multiple element additions—significant amounts of research and development have been conducted on the fabrication of these fuels. This paper presents an update of the US RERTR effort to develop fabrication techniques and the fabrication methods used for the RERTR-9A miniplate test.
Date: March 1, 2007
Creator: Clark, C.R.; Moore, G.A.; Jue, J.F.; Park, B.H.; Hallinan, N.P.; Wachs, D.M. et al.
Partner: UNT Libraries Government Documents Department

Analysis of the Reactor Physics of Low-Enrichment Fuel for the INL Advanced Test Reactor in support of RERTR

Description: Analysis of the performance of the ATR with a LEU fuel design shows promise in terms of a core design that will yield the same neutron sources in target locations. A proposed integral cladding burnable absorber design appears to meet power profile requirements that will satisfy power distributions for safety limits. Performance of this fuel design is ongoing; the current work is the initial evaluation of the core performance of this fuel design with increasing burnup. Results show that LEU fuel may have a longer lifetime that HEU fuel however, such limits may be set by mechanical performance of the fuel rather that available reactivity. Changes seen in the radial fuel power distribution with burnup in LEU fuel will require further study to ascertain the impact on neutron fluxes in target locations. Source terms for discharged fuel have also been studied. By its very nature, LEU fuel produces much more plutonium than is present in HEU fuel at discharge. However, the effect of the plutonium inventory appears to have little affect on radiotoxicity or decay heat in the fuel.
Date: June 1, 2012
Creator: DeHart, Mark; Skerjanc, William & Morrell, Sean
Partner: UNT Libraries Government Documents Department

Benchmark Evaluation of the Medium-Power Reactor Experiment Program Critical Configurations

Description: A series of small, compact critical assembly (SCCA) experiments were performed in 1962-1965 at the Oak Ridge National Laboratory Critical Experiments Facility (ORCEF) for the Medium-Power Reactor Experiment (MPRE) program. The MPRE was a stainless-steel clad, highly enriched uranium (HEU)-O2 fuelled, BeO reflected reactor design to provide electrical power to space vehicles. Cooling and heat transfer were to be achieved by boiling potassium in the reactor core and passing vapor directly through a turbine. Graphite- and beryllium-reflected assemblies were constructed at ORCEF to verify the critical mass, power distribution, and other reactor physics measurements needed to validate reactor calculations and reactor physics methods. The experimental series was broken into three parts, with the third portion of the experiments representing the beryllium-reflected measurements. The latter experiments are of interest for validating current reactor design efforts for a fission surface power reactor. The entire series has been evaluated as acceptable benchmark experiments and submitted for publication in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.
Date: February 1, 2013
Creator: Marshall, Margaret A. & Bess, John D.
Partner: UNT Libraries Government Documents Department

EVALUATION OF CORE PHYSICS ANALYSIS METHODS FOR CONVERSION OF THE INL ADVANCED TEST REACTOR TO LOW-ENRICHMENT FUEL

Description: Computational neutronics studies to support the possible conversion of the ATR to LEU are underway. Simultaneously, INL is engaged in a physics methods upgrade project to put into place modern computational neutronics tools for future support of ATR fuel cycle and experiment analysis. A number of experimental measurements have been performed in the ATRC in support of the methods upgrade project, and are being used to validate the new core physics methods. The current computational neutronics work is focused on performance of scoping calculations for the ATR core loaded with a candidate LEU fuel design. This will serve as independent confirmation of analyses that have been performed previously, and will evaluate some of the new computational methods for analysis of a candidate LEU fuel for ATR.
Date: April 1, 2012
Creator: DeHart, Mark & Chang, Gray S.
Partner: UNT Libraries Government Documents Department

Validation of FSP Reactor Design with Sensitivity Studies of Beryllium-Reflected Critical Assemblies

Description: The baseline design for space nuclear power is a fission surface power (FSP) system: sodium-potassium (NaK) cooled, fast spectrum reactor with highly-enriched-uranium (HEU)-O2 fuel, stainless steel (SS) cladding, and beryllium reflectors with B4C control drums. Previous studies were performed to evaluate modeling capabilities and quantify uncertainties and biases associated with analysis methods and nuclear data. Comparison of Zero Power Plutonium Reactor (ZPPR)-20 benchmark experiments with the FSP design indicated that further reduction of the total design model uncertainty requires the reduction in uncertainties pertaining to beryllium and uranium cross-section data. Further comparison with three beryllium-reflected HEU-metal benchmark experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) concluded the requirement that experimental validation data have similar cross section sensitivities to those found in the FSP design. A series of critical experiments was performed at ORCEF in the 1960s to support the Medium Power Reactor Experiment (MPRE) space reactor design. The small, compact critical assembly (SCCA) experiments were graphite- or beryllium-reflected assemblies of SS-clad, HEU-O2 fuel on a vertical lift machine. All five configurations were evaluated as benchmarks. Two of the five configurations were beryllium reflected, and further evaluated using the sensitivity and uncertainty analysis capabilities of SCALE 6.1. Validation of the example FSP design model was successful in reducing the primary uncertainty constituent, the Be(n,n) reaction, from 0.28 %dk/k to 0.0004 %dk/k. Further assessment of additional reactor physics measurements performed on the SCCA experiments may serve to further validate FSP design and operation.
Date: February 1, 2013
Creator: Bess, John D. & Marshall, Margaret A.
Partner: UNT Libraries Government Documents Department