395 Matching Results

Search Results

Advanced search parameters have been applied.

THE FINAL DEMISE OF EAST TENNESSEE TECHNOLOGY PARK BUILDING K-33 Health Physics Society Annual Meeting West Palm Beach, Florida June 27, 2011

Description: Building K-33 was constructed in 1954 as the final section of the five-stage uranium enrichment cascade at the Oak Ridge Gaseous Diffusion Plant (ORGDP). The two original building (K-25 and K-27) were used to produce weapons grade highly enriched uranium (HEU). Building K-29, K-31, and K-33 were added to produce low enriched uranium (LEU) for nuclear power plant fuel. During ORGDP operations K-33 produced a peak enrichment of 2.5%. Thousands of tons of reactor tails fed into gaseous diffusion plants in the 1950s and early 1960s introducing some fission products and transuranics. Building K-33 was a two-story, 25-meters (82-feet) tall structure with approximately 30 hectare (64 acres) of floor space. The Operations (first) Floor contained offices, change houses, feed vaporization rooms, and auxiliary equipment to support enrichment operations. The Cell (second) Floor contained the enrichment process equipment and was divided into eight process units (designated K-902-1 through K-902-8). Each unit contained ten cells, and each cell contained eight process stages (diffusers) for a total of 640 enrichment stages. 1985: LEU buildings were taken off-line after the anticipated demand for uranium enrichment failed to materialize. 1987: LEU buildings were placed in permanent shutdown. Process equipment were maintained in a shutdown state. 1997: DOE signed an Action Memorandum for equipment removal and decontamination of Buildings K-29, K-31, K-33; BNFL awarded contract to reindustrialize the buildings under the Three Buildings D&D and Recycle Project. 2002: Equipment removal complete and effort shifts to vacuuming, chemical cleaning, scabbling, etc. 2005: Decontamination efforts in K-33 cease. Building left with significant {sup 99}Tc contamination on metal structures and PCB contamination in concrete. Uranium, transuranics, and fission products also present on building shell. 2009: DOE targets Building K-33 for demolition. 2010: ORAU contracted to characterize Building K-33 for final disposition at the Environmental Management Waste Management Facility (EMWMF) ...
Date: June 27, 2011
Creator: King, David A.
Partner: UNT Libraries Government Documents Department

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

Revised FINAL–REPORT NO. 2: INDEPENDENT CONFIRMATORY SURVEY SUMMARY AND RESULTS FOR THE ENRICO FERMI ATOMIC POWER PLANT, UNIT 1, NEWPORT, MICHIGAN (DOCKET NO. 50 16; RFTA 10-004) 2018-SR-02-1

Description: The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor power with the first core loading (Core A) was 200 MW. The primary system was filled with sodium in December 1960 and criticality was achieved in August 1963. The reactor was tested at low power during the first couple years of operation. Power ascension testing above 1 MW commenced in December 1965 immediately following the receipt of a high-power operating license. In October 1966 during power ascension, zirconium plates at the bottom of the reactor vessel became loose and blocked sodium coolant flow to some fuel subassemblies. Two subassemblies started to melt and the reactor was manually shut down. No abnormal releases to the environment occurred. Forty-two months later after the cause had been determined, cleanup completed, and the fuel replaced, Fermi 1 was restarted. However, in November 1972, PRDC made the decision to decommission Fermi 1 as the core was approaching its burn-up limit. The fuel and blanket subassemblies were shipped off-site in 1973. Following that, the secondary sodium system was drained and sent off-site. The radioactive primary sodium was stored on-site in storage tanks and 55 gallon (gal) drums until it was shipped off-site in 1984. The initial decommissioning of Fermi 1 was completed in 1975. Effective January 23, 1976, DPR-9 was transferred to the Detroit Edison Company (DTE) as a 'possession only' license (DTE 2010a). This report ...
Date: October 27, 2011
Creator: Bailey, Erika
Partner: UNT Libraries Government Documents Department

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

FINAL REPORT –INDEPENDENT VERIFICATION SURVEY SUMMARY AND RESULTS FOR THE ARGONNE NATIONAL LABORATORY BUILDING 330 PROJECT FOOTPRINT, ARGONNE, ILLINOIS

Description: ORISE conducted onsite verification activities of the Building 330 project footprint during the period of June 6 through June 7, 2011. The verification activities included technical reviews of project documents, visual inspections, radiation surface scans, and sampling and analysis. The draft verification report was issued in July 2011 with findings and recommendations. The contractor performed additional evaluations and remediation.
Date: February 29, 2012
Creator: BAILEY, ERIKA N.
Partner: UNT Libraries Government Documents Department

Type A verification report for the high flux beam reactor stack and grounds, Brookhaven National Laboratory, Upton, New York

Description: The U.S. Department of Energy (DOE) Order 458.1 requires independent verification (IV) of DOE cleanup projects (DOE 2011). The Oak Ridge Institute for Science and Education (ORISE) has been designated as the responsible organization for IV of the High Flux Beam Reactor (HFBR) Stack and Grounds area at Brookhaven National Laboratory (BNL) in Upton, New York. The IV evaluation may consist of an in-process inspection with document and data reviews (Type A Verification) or a confirmatory survey of the site (Type B Verification). DOE and ORISE determined that a Type A verification of the documents and data for the HFBR Stack and Grounds: Survey Units (SU) 6, 7, and 8 was appropriate based on the initial survey unit classification, the walkover surveys, and the final analytical results provided by the Brookhaven Science Associates (BSA). The HFBR Stack and Grounds surveys began in June 2011 and were completed in September 2011. Survey activities by BSA included gamma walkover scans and sampling of the as-left soils in accordance with the BSA Work Procedure (BNL 2010a). The Field Sampling Plan - Stack and Remaining HFBR Outside Areas (FSP) stated that gamma walk-over surveys would be conducted with a bare sodium iodide (NaI) detector, and a collimated detector would be used to check areas with elevated count rates to locate the source of the high readings (BNL 2010b). BSA used the Mult- Agency Radiation Survey and Site Investigation Manual (MARSSIM) principles for determining the classifications of each survey unit. Therefore, SUs 6 and 7 were identified as Class 1 and SU 8 was deemed Class 2 (BNL 2010b). Gamma walkover surveys of SUs 6, 7, and 8 were completed using a 2�2 NaI detector coupled to a data-logger with a global positioning system (GPS). The 100% scan surveys conducted prior to the final status ...
Date: January 13, 2012
Creator: Harpenau, Evan M.
Partner: UNT Libraries Government Documents Department

FINAL–REPORT NO. 2: INDEPENDENT CONFIRMATORY SURVEY SUMMARY AND RESULTS FOR THE ENRICO FERMI ATOMIC POWER PLANT, UNIT 1, NEWPORT, MICHIGAN (DOCKET NO. 50 16; RFTA 10-004)

Description: The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor power with the first core loading (Core A) was 200 MW. The primary system was filled with sodium in December 1960 and criticality was achieved in August 1963.
Date: July 7, 2011
Creator: Bailey, Erika
Partner: UNT Libraries Government Documents Department

POTENTIAL BENCHMARKS FOR ACTINIDE PRODUCTION IN HANFORD REACTORS

Description: 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.
Date: October 19, 2011
Creator: RJ, PUIGH & H, TOFFER
Partner: UNT Libraries Government Documents Department

National Emission Standards for Hazardous Air Pollutants - Radionuclide Emissions, Calendar Year 2011

Description: The U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office operates the Nevada National Security Site (NNSS) and North Las Vegas Facility (NLVF). From 1951 through 1992, the NNSS was the continental testing location for U.S. nuclear weapons. The release of radionuclides from NNSS activities has been monitored since the initiation of atmospheric testing. Limitation to underground detonations after 1962 greatly reduced radiation exposure to the public surrounding the NNSS. After nuclear testing ended in 1992, NNSS radiation monitoring focused on detecting airborne radionuclides from historically contaminated soils. These radionuclides are derived from re-suspension of soil (primarily by wind) and emission of tritium-contaminated soil moisture through evapotranspiration. Low amounts of legacy-related tritium are also emitted to air at the NLVF, an NNSS support complex in North Las Vegas. To protect the public from harmful levels of man-made radiation, the Clean Air Act, National Emission Standards for Hazardous Air Pollutants (NESHAP) (Title 40 Code of Federal Regulations [CFR] Part 61 Subpart H) limits the release of radioactivity from a U.S. Department of Energy (DOE) facility to that which would cause 10 millirem per year (mrem/yr) effective dose equivalent to any member of the public. This limit does not include radiation unrelated to NNSS activities. Unrelated doses could come from naturally occurring radioactive elements, from sources such as medically or commercially used radionuclides, or from sources outside of the United States, such as the damaged Fukushima nuclear power plant in Japan. Radionuclides from the Fukushima nuclear power plant were detected at the NNSS in March 2011 and are discussed further in Section III. The NNSS demonstrates compliance with the NESHAP limit by using environmental measurements of radionuclide air concentrations at critical receptor locations. This method was approved by the EPA for use on the NNSS in 2001 and has been ...
Date: June 19, 2012
Creator: Monitoring, NSTec Ecological and Environmental
Partner: UNT Libraries Government Documents Department

Technical Letter Report: Evaluation and Analysis of a Few International Periodic Safety Review Summary Reports

Description: At the request of the United States (U.S.) government, the International Atomic Energy Agency (IAEA) assembled a team of 20 senior safety experts to review the regulatory framework for the safety of operating nuclear power plants in the United States. This review focused on the effectiveness of the regulatory functions implemented by the NRC and on its commitment to nuclear safety and continuous improvement. One suggestion resulting from that review was that the U.S. Nuclear Regulatory Commission (NRC) incorporate lessons learned from periodic safety reviews (PSRs) performed in other countries as an input to the NRC’s assessment processes. In the U.S., commercial nuclear power plants (NPPs) are granted an initial 40-year operating license, which may be renewed for additional 20-year periods, subject to complying with regulatory requirements. The NRC has established a framework through its inspection, and operational experience processes to ensure the safe operation of licensed nuclear facilities on an ongoing basis. In contrast, most other countries do not impose a specific time limit on the operating licenses for NPPs, they instead require that the utility operating the plant perform PSRs, typically at approximately 10-year intervals, to assure continued safe operation until the next assessment. The staff contracted with Argonne National Laboratory (Argonne) to perform a pilot review of selected translated PSR assessment reports and related documentation from foreign nuclear regulatory authorities to identify any potential new regulatory insights regarding license renewal-related topics and NPP operating experience (OpE). A total of 14 PSR assessment documents from 9 countries were reviewed. For all of the countries except France, individual reports were provided for each of the plants reviewed. In the case of France, three reports were provided that reviewed the performance assessment of thirty-four 900-MWe reactors of similar design commissioned between 1978 and 1988. All of the reports reviewed were ...
Date: December 17, 2013
Creator: Chopra, Omesh K.; Diercks, Dwight R.; Ma, David Chia-Chiun & Garud, Yogendra S.
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

Study of a multi-beam accelerator driven thorium reactor

Description: The primary advantages that accelerator driven systems have over critical reactors are: (1) Greater flexibility regarding the composition and placement of fissile, fertile, or fission product waste within the blanket surrounding the target, and (2) Potentially enhanced safety brought about by operating at a sufficiently low value of the multiplication factor to preclude reactivity induced events. The control of the power production can be achieved by vary the accelerator beam current. Furthermore, once the beam is shut off the system shuts down. The primary difference between the operation of an accelerator driven system and a critical system is the issue of beam interruptions of the accelerator. These beam interruptions impose thermo-mechanical loads on the fuel and mechanical components not found in critical systems. Studies have been performed to estimate an acceptable number of trips, and the value is significantly less stringent than had been previously estimated. The number of acceptable beam interruptions is a function of the length of the interruption and the mission of the system. Thus, for demonstration type systems and interruption durations of 1sec < t < 5mins, and t > 5mins 2500/yr and 50/yr are deemed acceptable. However, for industrial scale power generation without energy storage type systems and interruption durations of t < 1sec., 1sec < t < 10secs., 10secs < t < 5mins, and t > 5mins, the acceptable number of interruptions are 25000, 2500, 250, and 3 respectively. However, it has also been concluded that further development is required to reduce the number of trips. It is with this in mind that the following study was undertaken. The primary focus of this study will be the merit of a multi-beam target system, which allows for multiple spallation sources within the target/blanket assembly. In this manner it is possible to ameliorate the effects of ...
Date: March 1, 2011
Creator: Ludewig, H. & Aronson, A.
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

Thermochemical Modeling of the Uranium-Cerium-Oxygen System

Description: The objective of the Fuel Cycle R&D Program, Advanced Fuels campaign is to provide the research and development necessary to develop low loss, high quality nuclear fuels for ultra-high burnup reactor operation. Primary work in this area will be focused on the ceramic and metallic fuel systems. The goal of the current work is to enhance the understanding of ceramic nuclear fuel thermochemistry to support fuel research and development efforts. The thermochemical behavior of oxide nuclear fuel under irradiation is dependent on the oxygen to metal ratio (O:M). In fluorite-structured fuel, the actinide metal cation is bonded with {approx}2 oxygen atoms on a crystal lattice and as the metal atoms fission, fission fragments and free oxygen are created. The resulting fission fragments will contain some oxide forming elements, however these are insufficient to bind to all the liberated oxygen and therefore, there is an average increase in O:M with fuel burnup. Some of the fission products also form species that will migrate to and react with the cladding surface in a phenomenon known as Fuel Clad Chemical Interaction (FCCI). Cladding corrosion is life-limiting so it is desirable to understand influencing factors, such as oxide thermochemistry, which can be used to guide the design and fabrication of higher burn up fuel. A phased oxide fuel thermochemical model development effort is underway within the Advanced Fuels Campaign. First models of binary oxide systems are developed. For nuclear fuel system this means U-O and transuranic systems such as Pu-O, Np-O and Am-O. Next, the binary systems will be combined to form pseudobinary systems such as U-Pu-O, etc. The model development effort requires the use of data to allow optimization based on known thermochemical parameters as a function of composition and temperature. Available data is mined from the literature and supplemented by experimental work ...
Date: October 2010
Creator: Voit, Stewart L. & Besmann, Theodore M.
Partner: UNT Libraries Government Documents Department

Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

Description: This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be deployed commercially and have only been demonstrated in testing at a laboratory scale.
Date: October 1, 2011
Creator: Moses, David Lewis
Partner: UNT Libraries Government Documents Department

Fast Spectrum Molten Salt Reactor Options

Description: During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.
Date: July 1, 2011
Creator: Gehin, Jess C; Holcomb, David Eugene; Flanagan, George F; Patton, Bruce W; Howard, Rob L & Harrison, Thomas J
Partner: UNT Libraries Government Documents Department

Coated Particle and Deep Burn Fuels Monthly Highlights December 2010

Description: During FY 2011 the CP & DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for November 2010, ORNL/TM-2010/323, was distributed to program participants on December 9, 2010. The final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Thermochemical Data and Model Development - (a) Thermochemical Modeling, (b) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Pebble Bed Design (INL), (c) Radiation Damage and Properties; (2) TRISO (tri-structural isotropic) Development - (a) TRU (transuranic elements) Kernel Development, (b) Coating Development; (3) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing (ORNL); (4) Fuel Performance and Analytical Analysis - Fuel Performance Modeling (ORNL).
Date: January 1, 2011
Creator: Snead, Lance Lewis; Bell, Gary L & Besmann, Theodore M
Partner: UNT Libraries Government Documents Department

Compatibility/Stability Issues in the Use of Nitride Kernels in LWR TRISO Fuel

Description: The stability of the SiC layer in the presence of free nitrogen will be dependent upon the operating temperatures and resulting nitrogen pressures whether it is at High Temperature Gas-Cooled Reactor (HTGR) temperatures of 1000-1400 C (coolant design dependent) or LWR temperatures that range from 500-700 C. Although nitrogen released in fissioning will form fission product nitrides, there will remain an overpressure of nitrogen of some magnitude. The nitrogen can be speculated to transport through the inner pyrolytic carbon layer and contact the SiC layer. The SiC layer may be envisioned to fail due to resulting nitridation at the elevated temperatures. However, it is believed that these issues are particularly avoided in the LWR application. Lower temperatures will result in significantly lower nitrogen pressures. Lower temperatures will also substantially reduce nitrogen diffusion rates through the layers and nitriding kinetics. Kinetics calculations were performed using an expression for nitriding silicon. In order to further address these concerns, experiments were run with surrogate fuel particles under simulated operating conditions to determine the resulting phase formation at 700 and 1400 C.
Date: February 1, 2012
Creator: Armstrong, Beth L & Besmann, Theodore M
Partner: UNT Libraries Government Documents Department

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights October 2010

Description: The DB Program monthly highlights report for September 2010, ORNL/TM-2010/252, was distributed to program participants by email on October 26. This report discusses: (1) Core and Fuel Analysis; (2) Spent Fuel Management; (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor); (4) TRU (transuranic elements) HTR Fuel Qualification; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle.
Date: November 1, 2010
Creator: Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D & Bell, Gary L
Partner: UNT Libraries Government Documents Department

Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010

Description: The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).
Date: October 1, 2010
Creator: Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D & Bell, Gary L
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

Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration

Description: Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor deployment analysis and small grid stability issues. Objective 2 - Develop and test an autonomous and fault-tolerant control architecture and apply to the IRIS system and an experimental flow control loop, with extensions to multiple reactor modules, nuclear desalination, and optimal sensor placement strategy. Objective 3 - Develop and test an integrated monitoring, diagnosis, and prognosis system for SMRs using the IRIS as a test platform, and integrate process and equipment monitoring (PEM) and process and equipment prognostics (PEP) toolboxes. The research tasks are focused on meeting the unique needs of reactors that may be deployed to remote locations or to developing countries with limited support infrastructure. These applications will require smaller, robust reactor designs with advanced technologies for sensors, instrumentation, and control. An excellent overview of SMRs is described in an article by Ingersoll (2009). The article refers to these as deliberately small reactors. Most of these have modular characteristics, with multiple units deployed at the same plant site. Additionally, the topics focus on meeting two of the eight needs outlined in the recently published 'Technology Roadmap on Instrumentation, ...
Date: May 31, 2011
Creator: Hines, J. Wesley; Upadhyaya, Belle R.; Doster, J. Michael; Edwards, Robert M.; Lewis, Kenneth D.; Turinsky, Paul et al.
Partner: UNT Libraries Government Documents Department

Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

Description: 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. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.
Date: April 4, 2012
Creator: Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division) et al.
Partner: UNT Libraries Government Documents Department