National Library of Energy BETA

Sample records for reactors generation spent

  1. Spent nuclear fuel discharges from U.S. reactors 1994

    SciTech Connect (OSTI)

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  2. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect (OSTI)

    Not Available

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  3. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect (OSTI)

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  4. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    DOE Patents [OSTI]

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  5. Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    National Nuclear Security Administration (NNSA)

    * Complete reactor control rod system. * Note: Does not include the steam turbine generator portion of the power plant. - Sensitive nuclear technology: Any information...

  6. A Compact Reactor Gate Discharge Monitor for Spent Fuel.

    SciTech Connect (OSTI)

    Franco, J. B.; Menlove, Howard O.; Eccleston, G. W.; Miller, M. C.

    2005-01-01

    This paper presents a new design for a reactor gate discharge monitor that has evolved from the baseline discharge monitors used at the Fugen and Tokai-1 reactors in Japan. The main design innovation is the ability to determine direction-of-motion of spent fuel using a single sensor module, as opposed to the two modules used in both baseline design systems. Use of a single module reduces the final system complexity and weight significantly without compromising functionality. The reactor gate discharge monitor uses standard International Atomic Energy Agency (IAEA) hardware and software components. The requirements to determine direction-of-motion from a single module precipitated several development efforts described herein in both the MiniGRAND data acquisition hardware and in the uninterruptible power supply source.

  7. EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of a proposal to accept 409 spent fuel elements from eight foreign research reactors in seven European countries.  The spent fuel would be shipped across...

  8. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect (OSTI)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactorsspent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  9. Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors

    Energy Savers [EERE]

    6 Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Reactor Sites December 2008 U.S. Department of Energy Office of Civilian Radioactive Waste Management Washington, D.C. Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel The picture on the cover is the Connecticut Yankee Independent Spent Fuel Storage Installation site in Haddam, Connecticut, with 43 dry storage NRC-licensed dual-purpose

  10. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    SciTech Connect (OSTI)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2 discharge reuse. The EM2 waste disposal profile is effectively only fission products, which reduces the mass (about 3% vs LWR), average half life, heat and long term radio-toxicity of the disposal. Widespread implementation of EM2 fuel cycle is highly significant as it would increase world energy reserves; the remaining energy in U.S. LWR SNF alone exceeds that in the U.S. natural gas reserves. Unlike many LWR SNF disposition concepts, the EM2 fuel cycle conversion of SNF produces energy and associated revenue such that the overall project is cost effective. By providing conversion of SNF to fission products the fuel cycle is closed and a non-repository LWR SNF disposition path is created and overall repository requirements are significantly reduced. (authors)

  11. Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

    SciTech Connect (OSTI)

    Hellesen, C.; Grape, S.; Haakanson, A.; Jacobson Svaerd, S.; Jansson, P.

    2013-07-01

    Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

  12. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    SciTech Connect (OSTI)

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  13. The burnup dependence of light water reactor spent fuel oxidation

    SciTech Connect (OSTI)

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5).

  14. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    SciTech Connect (OSTI)

    Gauld, Ian C; Sly, Nicholas C; Michel-Sendis, Franco

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  15. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  16. Electricity Generating Portfolios with Small Modular Reactors

    Broader source: Energy.gov [DOE]

    A paper by Geoffrey Rothwell, Ph.D., Stanford University (retired), and Francesco Ganda, Ph.D., Argonne National Laboratory on "Electricity Generating Portfolios with Small Modular Reactors".

  17. DOE/EIS-0218-SA-3: Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program (November 2004)

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM NOVEMBER 2004 DOE/EIS-0218-SA-3 U.S. Department of Energy National Nuclear Security Administration Washington, DC Final Supplement Analysis for the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program Final i TABLE OF CONTENTS Page 1. Introduction.............................................................................................................................................. 1 2.

  18. Thermal analysis for a spent reactor fuel storage test in granite

    SciTech Connect (OSTI)

    Montan, D.N.

    1980-09-01

    A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy`s Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m{sup 3}/s and an ambient rock temperature of 23{sup 0}C, the maximum calculated rock temperature (near the center of the heat source) is about 100{sup 0}C while the maximum air temperature in the drift is around 40{sup 0}C. This ventilation (1 m{sup 3}/s through the main drift and 1/2 m{sup 3}/s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage.

  19. Spent nuclear fuel discharges from US reactors 1992

    SciTech Connect (OSTI)

    Not Available

    1994-05-05

    This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

  20. Monte Carlo analysis of neutron slowing-down-time spectrometer for fast reactor spent fuel assay

    SciTech Connect (OSTI)

    Chen, Jianwei; Lineberry, Michael

    2007-07-01

    Using the neutron slowing-down-time method as a nondestructive assay tool to improve input material accountancy for fast reactor spent fuel reprocessing is under investigation at Idaho State University. Monte Carlo analyses were performed to simulate the neutron slowing down process in different slowing down spectrometers, namely, lead and graphite, and determine their main parameters. {sup 238}U threshold fission chamber response was simulated in the Monte Carlo model to represent the spent fuel assay signals, the signature (fission/time) signals of {sup 235}U, {sup 239}Pu, and {sup 241}Pu were simulated as a convolution of fission cross sections and neutron flux inside the spent fuel. {sup 238}U detector signals were analyzed using linear regression model based on the signatures of fissile materials in the spent fuel to determine weight fractions of fissile materials in the Advanced Burner Test Reactor spent fuel. The preliminary results show even though lead spectrometer showed a better assay performance than graphite, graphite spectrometer could accurately determine weight fractions of {sup 239}Pu and {sup 241}Pu given proper assay energy range were chosen. (authors)

  1. Waste Generated from LMR-AMTEC Reactor Concept

    SciTech Connect (OSTI)

    Hasan, Ahmed; Mohamed, Yasser, T.; Mohammaden, Tarek, F.

    2003-02-25

    The candidate Liquid Metal Reactor-Alkali Metal Thermal -to- Electric Converter (LMR-AMTEC) is considered to be the first reactor that would use pure liquid potassium as a secondary coolant, in which potassium vapor aids in the conversion of thermal energy to electric energy. As with all energy production, the thermal generation of electricity produces wastes. These wastes must be managed in ways which safeguard human health and minimize their impact on the environment. Nuclear power is the only energy industry, which takes full responsibility for all its wastes. Based on the candidate design of the LMR-AMTEC components and the coolant types, different wastes will be generated from LMR. These wastes must be classified and characterized according to the U.S. Code of Federal Regulation, CFR. This paper defines the waste generation and waste characterization from LMR-AMTEC and reviews the applicable U.S. regulations that govern waste transportation, treatment, storage and final disposition. The wastes generated from LMR-AMTEC are characterized as: (1) mixed waste which is generated from liquid sodium contaminated by fission products and activated corrosion products; (2) hazardous waste which is generated from liquid potassium contaminated by corrosion products; (3) spent nuclear fuel; and (4) low-level radioactive waste which is generated from the packing materials (e.g. activated carbon in cold trap and purification units). The regulations and management of these wastes are summarized in this paper.

  2. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    SciTech Connect (OSTI)

    Pope, R B; Diggs, J M [eds.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  3. Second generation Research Reactor Fuel Container (RRFC-II).

    SciTech Connect (OSTI)

    Abhold, M. E.; Baker, M. C.; Bourret, S. C.; Harker, W. C.; Pelowitz, D. G.; Polk, P. J.

    2001-01-01

    The second generation Research Reactor Fuel Counter (RRFC-II) has been developed to measure the remaining {sup 235}U content in foreign spent Material Test Reactor (MTR)-type fuel being returned to the Westinghouse Savannah River Site (WSRS) for interim storage and subsequent disposal. The fuel to be measured started as fresh fuel nominally with 93% enriched Uraniuin alloyed with A1 clad in Al. The fuel was irradiated to levels of up to 65% burnup. The RRFC-II, which will be located in the L-Basin spent fuel pool, is intended to assay the {sup 235}U content using a combination of passive neutron coincidence counting, active neutron coincidence counting, and active-multiplicity analysis. Measurements will be done underwater, eliminating the need for costly and hazardous handling operations of spent fuel out of water. The underwater portion of the RRFC-II consists of a watertight stainless steel housing containing neutron and gamma detectors and a scanning active neutron source. The portion of the system that resides above water consists of data-processing electronics; electromechanical drive electronics; a computer to control the operation of the counter, to collect, and to analyze data; and a touch screen interface located at the equipment rack. The RRFC-II is an improved version of the Los Alamos-designed RRFC already installed in the SRS Receipts Basin for Offsite Fuel. The RRFC-II has been fabricated and is scheduled for installation in late FY 2001 pending acceptance testing by Savannah River Site personnel.

  4. Microstructural characteristics of PWR [pressurized water reactor] spent fuel relative to its leaching behavior

    SciTech Connect (OSTI)

    Wilson, C.N.

    1986-01-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

  5. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    SciTech Connect (OSTI)

    Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  6. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    SciTech Connect (OSTI)

    Not Available

    1992-07-01

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

  7. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect (OSTI)

    Ilas, Germina; Gauld, Ian C

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  8. Plasma generators, reactor systems and related methods

    DOE Patents [OSTI]

    Kong, Peter C. (Idaho Falls, ID); Pink, Robert J. (Pocatello, ID); Lee, James E. (Idaho Falls, ID)

    2007-06-19

    A plasma generator, reactor and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably coupled together to form a chamber. Associated with each section is an electrode set including three electrodes with each electrode being coupled to a single phase of a three-phase alternating current (AC) power supply. The electrodes are disposed about a longitudinal centerline of the chamber and are arranged to provide and extended arc and generate an extended body of plasma. The electrodes are displaceable relative to the longitudinal centerline of the chamber. A control system may be utilized so as to automatically displace the electrodes and define an electrode gap responsive to measure voltage or current levels of the associated power supply.

  9. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  10. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, M.M.; Lau, L.K.; Schulz, T.L.

    1993-12-14

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

  11. RGG: Reactor geometry (and mesh) generator

    SciTech Connect (OSTI)

    Jain, R.; Tautges, T.

    2012-07-01

    The reactor geometry (and mesh) generator RGG takes advantage of information about repeated structures in both assembly and core lattices to simplify the creation of geometry and mesh. It is released as open source software as a part of the MeshKit mesh generation library. The methodology operates in three stages. First, assembly geometry models of various types are generated by a tool called AssyGen. Next, the assembly model or models are meshed by using MeshKit tools or the CUBIT mesh generation tool-kit, optionally based on a journal file output by AssyGen. After one or more assembly model meshes have been constructed, a tool called CoreGen uses a copy/move/merge process to arrange the model meshes into a core model. In this paper, we present the current state of tools and new features in RGG. We also discuss the parallel-enabled CoreGen, which in several cases achieves super-linear speedups since the problems fit in available RAM at higher processor counts. Several RGG applications - 1/6 VHTR model, 1/4 PWR reactor core, and a full-core model for Monju - are reported. (authors)

  12. Electricity Generating Portfolios with Small Modular Reactors | Department

    Office of Environmental Management (EM)

    of Energy Electricity Generating Portfolios with Small Modular Reactors Electricity Generating Portfolios with Small Modular Reactors This paper provides a method for estimating the probability distributions of the levellized costs of electricity. These distributions can be used to find cost-risk minimizing portfolios of electricity generating assets including Combined-Cycle Gas Turbines, coal-fired power plants with sulfur scrubbers, and Small Modular Reactors, SMRs. PDF icon Electricity

  13. Plasma generators, reactor systems and related methods - Energy...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Visit the Technology Transfer and Commercialization Office Website Abstract: A plasma generator, reactor and associated systems and methods are provided in accordance with the...

  14. Generating unstructured nuclear reactor core meshes in parallel

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Jain, Rajeev; Tautges, Timothy J.

    2014-10-24

    Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

  15. Generating unstructured nuclear reactor core meshes in parallel

    SciTech Connect (OSTI)

    Jain, Rajeev; Tautges, Timothy J.

    2014-10-24

    Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor core examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.

  16. Advantages of co-located spent fuel reprocessing, repository and underground reactor facilities

    SciTech Connect (OSTI)

    Mahar, James M.; Kunze, Jay F.; Wes Myers, Carl; Loveland, Ryan

    2007-07-01

    The purpose of this work is to extend the discussion of potential advantages of the underground nuclear park (UNP) concept by making specific concept design and cost estimate comparisons for both present Generation III types of reactors and for some of the modular Gen IV or the GNEP modular concept. For the present Gen III types, we propose co-locating reprocessing and (re)fabrication facilities along with disposal facilities in the underground park. The goal is to determine the site costs and facility construction costs of such a complex which incorporates the advantages of a closed fuel cycle, nuclear waste repository, and ultimate decommissioning activities all within the UNP. Modular power generation units are also well-suited for placement underground and have the added advantage of construction using current and future tunnel boring machine technology. (authors)

  17. Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

    SciTech Connect (OSTI)

    S. D. Herrmann; S. X. Li

    2010-09-01

    A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl – 1 wt% Li2O at 650 °C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 °C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

  18. Extended Storage for Research and Test Reactor Spent Fuel for 2006 and Beyond

    SciTech Connect (OSTI)

    Hurt, William Lon; Moore, K.M.; Shaber, Eric Lee; Mizia, Ronald Eugene

    1999-10-01

    This paper will examine issues associated with extended storage of a variety of spent nuclear fuels. Recent experiences at the Idaho National Engineering and Environmental Laboratory and Hanford sites will be described. Particular attention will be given to storage of damaged or degraded fuel. The first section will address a survey of corrosion experience regarding wet storage of spent nuclear fuel. The second section will examine issues associated with movement from wet to dry storage. This paper also examines technology development needs to support storage and ultimate disposition.

  19. Process and apparatus for generating elemental sulfur and re-usable metal oxide from spent metal sulfide sorbents

    DOE Patents [OSTI]

    Ayala, Raul E. (Clifton Park, NY); Gal, Eli (Lititz, PA)

    1995-01-01

    A process and apparatus for generating elemental sulfur and re-usable metal oxide from spent metal-sulfur compound. Spent metal-sulfur compound is regenerated to re-usable metal oxide by moving a bed of spent metal-sulfur compound progressively through a single regeneration vessel having a first and second regeneration stage and a third cooling and purging stage. The regeneration is carried out and elemental sulfur is generated in the first stage by introducing a first gas of sulfur dioxide which contains oxygen at a concentration less than the stoichiometric amount required for complete oxidation of the spent metal-sulfur compound. A second gas containing sulfur dioxide and excess oxygen at a concentration sufficient for complete oxidation of the partially spent metal-sulfur compound, is introduced into the second regeneration stage. Gaseous sulfur formed in the first regeneration stage is removed prior to introducing the second gas into the second regeneration stage. An oxygen-containing gas is introduced into the third cooling and purging stage. Except for the gaseous sulfur removed from the first stage, the combined gases derived from the regeneration stages which are generally rich in sulfur dioxide and lean in oxygen, are removed from the regenerator as an off-gas and recycled as the first and second gas into the regenerator. Oxygen concentration is controlled by adding air, oxygen-enriched air or pure oxygen to the recycled off-gas.

  20. Process and apparatus for recovery of fissionable materials from spent reactor fuel by anodic dissolution

    DOE Patents [OSTI]

    Tomczuk, Zygmunt (Orland Park, IL); Miller, William E. (Naperville, IL); Wolson, Raymond D. (Lockport, IL); Gay, Eddie C. (Park Forest, IL)

    1991-01-01

    An electrochemical process and apparatus for the recovery of uranium and plutonium from spent metal clad fuel pins is disclosed. The process uses secondary reactions between U.sup.+4 cations and elemental uranium at the anode to increase reaction rates and improve anodic efficiency compared to prior art processes. In another embodiment of the process, secondary reactions between Cd.sup.+2 cations and elemental uranium to form uranium cations and elemental cadmium also assists in oxidizing the uranium at the anode.

  1. Steam generator for liquid metal fast breeder reactor

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Garner, Daniel C. (Murrysville, PA); Wineman, Arthur L. (Greensburg, PA); Robey, Robert M. (North Huntingdon, PA)

    1985-01-01

    Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

  2. AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR

    SciTech Connect (OSTI)

    Dewes, J.

    2014-02-24

    The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/C was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.

  3. US Spent (Used) Fuel Status, Management and Likely Directions- 12522

    SciTech Connect (OSTI)

    Jardine, Leslie J.

    2012-07-01

    As of 2010, the US has accumulated 65,200 MTU (42,300 MTU of PWR's; 23,000 MTU of BWR's) of spent (irradiated or used) fuel from 104 operating commercial nuclear power plants situated at 65 sites in 31 States and from previously shutdown commercial nuclear power plants. Further, the Department of Energy (DOE) has responsibility for an additional 2458 MTU of DOE-owned defense and non defense spent fuel from naval nuclear power reactors, various non-commercial test reactors and reactor demonstrations. The US has no centralized large spent fuel storage facility for either commercial spent fuel or DOE-owned spent fuel. The 65,200 MTU of US spent fuel is being safely stored by US utilities at numerous reactor sites in (wet) pools or (dry) metal or concrete casks. As of November 2010, the US had 63 'independent spent fuel storage installations' (or ISFSI's) licensed by the US Nuclear Regulatory Commission located at 57 sites in 33 states. Over 1400 casks loaded with spent fuel for dry storage are at these licensed ISFSI's; 47 sites are located at commercial reactor sites and 10 are located 'away' from a reactor (AFR's) site. DOE's small fraction of a 2458 MTU spent fuel inventory, which is not commercial spent fuel, is with the exception of 2 MTU, being stored at 4 sites in 4 States. The decades old US policy of a 'once through' fuel cycle with no recycle of spent fuel was set into a state of 'mass confusion or disruption' when the new US President Obama's administration started in early 2010 stopping the only US geologic disposal repository at the Yucca Mountain site in the State of Nevada from being developed and licensed. The practical result is that US nuclear power plant operators will have to continue to be responsible for managing and storing their own spent fuel for an indefinite period of time at many different sites in order to continue to generate electricity because there is no current US government plan, schedule or policy for taking possession of accumulated spent fuel from the utilities. There are technical solutions for continuing the safe storage of spent fuel for 100 years or more and these solutions will be implemented by the US utilities that need to keep their nuclear power plants operating while the unknown political events are played out to establish future US policy decisions that can remain in place long enough regarding accumulated spent fuel inventories to implement any new US spent fuel centralized storage or disposition policy by the US government. (author)

  4. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Jiang, Hao

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10? cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum of tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m?¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were from 1.22 10? to 4.70 10?, when the amplitude varied from 15.2 to 7.6 N·m. The measurements at the interrupts indicated a range of flexural rigidity from 30 to 50 Nm². The on-line monitoring revealed that the flexural rigidity was a little lower due to the high level of loading, from 25 to 42 Nm². Generally, no substantial change of rigidity was observed based on on-line monitoring during the cyclic fatigue testing process. Overall, the decreasing trend of lifetime with the increasing amplitude is well defined.

  5. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  6. Spent Nuclear Fuel

    Gasoline and Diesel Fuel Update (EIA)

    Nuclear & Uranium Glossary › FAQS › Overview Data Status of U.S. Nuclear Outages (interactive) Summary Uranium & nuclear fuel Nuclear power plants Spent nuclear fuel International All nuclear data reports Analysis & Projections Major Topics Most popular Nuclear plants and reactors Projections Recurring Uranium All reports Browse by Tag Alphabetical Frequency Tag Cloud Previous releases 2002 1998 Spent Nuclear Fuel Release date: December 7, 2015 Next release date: Late 2018 Spent

  7. State of Nevada comments on the OCRWM from-reactor spent fuel shipping cask preliminary design reports

    SciTech Connect (OSTI)

    Halstead, R.J.; Audin, L.; Hoskins, R.E.; Snedeker, D.F.

    1990-12-01

    The design of spent fuels shipping casks is described. Two casks from two different contractors are presented. The design needs are based on the OCRWM'S program specifications. (CBS)

  8. Small Modular Reactors - Key to Future Nuclear Power Generation in the U.S.

    Energy Savers [EERE]

    | Department of Energy Reactors - Key to Future Nuclear Power Generation in the U.S. Small Modular Reactors - Key to Future Nuclear Power Generation in the U.S. This report presents the results of an extensive analysis of the economics of both gigawatt-scale and small modular reactors. Topics covered include the safety case, economics, the business case, and a business plan, government incentives, licensing, design and engineering, and future research. PDF icon Small Modular Reactors - Key

  9. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    SciTech Connect (OSTI)

    Corwin, William R; Burchell, Timothy D; Katoh, Yutai; McGreevy, Timothy E; Nanstad, Randy K; Ren, Weiju; Snead, Lance Lewis; Wilson, Dane F

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOE's structural materials research activities being conducted to support VHTR development. By far, the largest portion of material's R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water reactor steels for anticipated VHTR off-normal conditions must be determined, as well as the effects of aging on tensile, creep, and toughness properties, and on thermal emissivity. (b) Large-scale fabrication process for higher temperature alloys, such as 9Cr-1MoV, including ensuring thick-section and weldment integrity must be developed, as well as improved definitions of creep-fatigue and negligible creep behavior. (5) High-Temperature Alloys: (a) Qualification and codification of materials for the intermediate heat exchanger, such as Alloys 617 or 230, for long-term very high-temperature creep, creep-fatigue, and environmental aging degradation must be done, especially in thin sections for compact designs, for both base metal and weldments. (b) Constitutive models and an improved methodology for high-temperature design must be developed.

  10. State of Nevada comments on the OCRWM from-reactor spent fuel shipping cask preliminary design reports

    SciTech Connect (OSTI)

    Halstead, R.J.; Audin, L.; Hoskins, R.E.; Snedeker, D.F.

    1990-12-01

    The design of spent fuels shipping casks is described. Two casks from two different contractors are presented. The design needs are based on the OCRWM`S program specifications. (CBS)

  11. Selection of materials for sodium fast reactor steam generators

    SciTech Connect (OSTI)

    Dubiez-Le Goff, S.; Garnier, S.; Gelineau, O.; Dalle, F.; Blat-Yrieix, M.; Augem, J. M.

    2012-07-01

    Sodium Fast Reactor (SFR) is considered in France as the most mature technology of the different Generation IV systems. In the short-term the designing work is focused on the identification of the potential tracks to demonstrate licensing capability, availability, in-service inspection capability and economical performance. In that frame materials selection for the major components, as the steam generator, is a particularly key point managed within a French Research and Development program launched by AREVA, CEA and EDF. The choice of the material for the steam generator is indeed complex because various aspects shall be considered like mechanical and thermal properties at high temperature, interaction with sodium on one side and water and steam on the other side, resistance to wastage, procurement, fabrication, weldability and ability for inspection and in-situ intervention. The following relevant options are evaluated: the modified 9Cr1Mo ferritic-martensitic grade and the Alloy 800 austenitic grade. The objective of this paper is to assess for both candidates their abilities to reach the current SFR needs regarding material design data, from AFCEN RCC-MRx Code in particular, compatibility with environments and manufacturability. (authors)

  12. Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 3, Site team reports

    SciTech Connect (OSTI)

    Not Available

    1993-11-01

    A self assessment was conducted of those Hanford facilities that are utilized to store Reactor Irradiated Nuclear Material, (RINM). The objective of the assessment is to identify the Hanford inventories of RINM and the ES & H concerns associated with such storage. The assessment was performed as proscribed by the Project Plan issued by the DOE Spent Fuel Working Group. The Project Plan is the plan of execution intended to complete the Secretary`s request for information relevant to the inventories and vulnerabilities of DOE storage of spent nuclear fuel. The Hanford RINM inventory, the facilities involved and the nature of the fuel stored are summarized. This table succinctly reveals the variety of the Hanford facilities involved, the variety of the types of RINM involved, and the wide range of the quantities of material involved in Hanford`s RINM storage circumstances. ES & H concerns are defined as those circumstances that have the potential, now or in the future, to lead to a criticality event, to a worker radiation exposure event, to an environmental release event, or to public announcements of such circumstances and the sensationalized reporting of the inherent risks.

  13. Plasma generators, reactor systems and related methods - Energy...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably...

  14. Long-term management of high-level radioactive waste (HLW) and spent

    Energy Savers [EERE]

    nuclear fuel (SNF) | Department of Energy Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor to generate nuclear energy but that has been removed from the reactor as no

  15. American National Standard: design requirements for light water reactor spent fuel storage facilities at nuclear power plants

    SciTech Connect (OSTI)

    Not Available

    1983-10-07

    This standard presents necessary design requirements for facilities at nuclear power plants for the storage and preparation for shipment of spent fuel from light-water moderated and cooled nuclear power stations. It contains requirements for the design of fuel storage pool; fuel storage racks; pool makeup, instrumentation and cleanup systems; pool structure and integrity; radiation shielding; residual heat removal; ventilation, filtration and radiation monitoring systems; shipping cask handling and decontamination; building structure and integrity; and fire protection and communication.

  16. Loss of spent fuel pool cooling PRA: Model and results

    SciTech Connect (OSTI)

    Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

    1996-09-01

    This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 {times} 10{sup {minus}5} and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 {times} 10{sup {minus}3}. Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible.

  17. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOE Patents [OSTI]

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  18. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOE Patents [OSTI]

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  19. 324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

    SciTech Connect (OSTI)

    HUMPHREYS, D C

    2002-08-01

    A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Team counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.

  20. Selection of a suitable reactor type for water desalination and power generation in Saudi Arabia

    SciTech Connect (OSTI)

    Hussein, F.M.

    1988-03-01

    Selection of a reactor type suitable for water desalination and power generation is a complex process that involves the evaluation of many criteria and requires the professional judgment of many experts in different fields. A reactor type that is suitable for one country might not be suitable for another. This is especially true in the case of Saudi Arabia because of its strategic location, the nature of its land and people, and its moderate technological situation. A detailed study using a computer code based on Saaty's mathematical pairwise comparison technique and developed in a previous study was carried out to find the most suitable reactor for water desalination and power generation in Saudi Arabia from among five potential types: boiling water reactors (BWRs), pressurized water reactors, CANDU heavy water reactors (HWRs), steam-generating heavy water reactors (SGHWRs), and high-temperature gas-cooled reactors. It was concluded that the CANDU HWR is the most suitable type for this purpose followed first by the BWR, then the SGHWR.

  1. Five Years of Building the Next Generation of Reactors | Department of

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Energy Years of Building the Next Generation of Reactors Five Years of Building the Next Generation of Reactors August 15, 2012 - 5:17pm Addthis Simulated three-dimensional fission power distribution of a single 17x17 rod PWR fuel assembly. | Photo courtesy of the Consortium for Advanced Simulation of Light Water Reactors (CASL). Simulated three-dimensional fission power distribution of a single 17x17 rod PWR fuel assembly. | Photo courtesy of the Consortium for Advanced Simulation of Light

  2. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    SciTech Connect (OSTI)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  3. Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

    SciTech Connect (OSTI)

    Corwin, William R; Burchell, Timothy D; Halsey, William; Hayner, George; Katoh, Yutai; Klett, James William; McGreevy, Timothy E; Nanstad, Randy K; Ren, Weiju; Snead, Lance Lewis; Stoller, Roger E; Wilson, Dane F

    2005-12-01

    The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor - SCWR and the Very High Temperature Reactor - VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor - GFR, the Lead-cooled Fast Reactor - LFR, and the Sodium-cooled Fast Reactor - SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides, and may provide an alternative to accelerator-driven systems. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development.

  4. Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 2, Working Group Assessment Team reports; Vulnerability development forms; Working group documents

    SciTech Connect (OSTI)

    Not Available

    1993-11-01

    The Secretary of Energy`s memorandum of August 19, 1993, established an initiative for a Department-wide assessment of the vulnerabilities of stored spent nuclear fuel and other reactor irradiated nuclear materials. A Project Plan to accomplish this study was issued on September 20, 1993 by US Department of Energy, Office of Environment, Health and Safety (EH) which established responsibilities for personnel essential to the study. The DOE Spent Fuel Working Group, which was formed for this purpose and produced the Project Plan, will manage the assessment and produce a report for the Secretary by November 20, 1993. This report was prepared by the Working Group Assessment Team assigned to the Hanford Site facilities. Results contained in this report will be reviewed, along with similar reports from all other selected DOE storage sites, by a working group review panel which will assemble the final summary report to the Secretary on spent nuclear fuel storage inventory and vulnerability.

  5. Benchmark Development in Support of Generation-IV Reactor Validation (IRPhEP 2010 Handbook)

    SciTech Connect (OSTI)

    John D. Bess; J. Blair Briggs

    2010-06-01

    The March 2010 edition of the International Reactor Physics Experiment Evaluation Project (IRPhEP) Handbook includes additional benchmark data that can be implemented in the validation of data and methods for Generation IV (GEN-IV) reactor designs. Evaluations supporting sodium-cooled fast reactor (SFR) efforts include the initial isothermal tests of the Fast Flux Test Facility (FFTF) at the Hanford Site, the Zero Power Physics Reactor (ZPPR) 10B and 10C experiments at the Idaho National Laboratory (INL), and the burn-up reactivity coefficient of Japan’s JOYO reactor. An assessment of Russia’s BFS-61 assemblies at the Institute of Physics and Power Engineering (IPPE) provides additional information for lead-cooled fast reactor (LFR) systems. Benchmarks in support of the very high temperature reactor (VHTR) project include evaluations of the HTR-PROTEUS experiments performed at the Paul Scherrer Institut (PSI) in Switzerland and the start-up core physics tests of Japan’s High Temperature Engineering Test Reactor. The critical configuration of the Power Burst Facility (PBF) at the INL which used ternary ceramic fuel, U(18)O2-CaO-ZrO2, is of interest for fuel cycle research and development (FCR&D) and has some similarities to “inert-matrix” fuels that are of interest in GEN-IV advanced reactor design. Two additional evaluations were revised to include additional evaluated experimental data, in support of light water reactor (LWR) and heavy water reactor (HWR) research; these include reactor physics experiments at Brazil’s IPEN/MB-01 Research Reactor Facility and the French High Flux Reactor (RHF), respectively. The IRPhEP Handbook now includes data from 45 experimental series (representing 24 reactor facilities) and represents contributions from 15 countries. These experimental measurements represent large investments of infrastructure, experience, and cost that have been evaluated and preserved as benchmarks for the validation of methods and collection of data in support of current and future reactor design and development.

  6. Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation

    SciTech Connect (OSTI)

    Gauld, Ian C.; Giaquinto, J. M.; Delashmitt, J. S.; Hu, Jianwei; Ilas, Germina; Haverlock, T. J.; Romano, Catherine E.

    2016-01-01

    Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via high performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.

  7. Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Gauld, Ian C.; Giaquinto, J. M.; Delashmitt, J. S.; Hu, Jianwei; Ilas, Germina; Haverlock, T. J.; Romano, Catherine E.

    2016-01-01

    Destructive radiochemical assay measurements of spent nuclear fuel rod segments from an assembly irradiated in the Three Mile Island unit 1 (TMI-1) pressurized water reactor have been performed at Oak Ridge National Laboratory (ORNL). Assay data are reported for five samples from two fuel rods of the same assembly. The TMI-1 assembly was a 15 X 15 design with an initial enrichment of 4.013 wt% 235U, and the measured samples achieved burnups between 45.5 and 54.5 gigawatt days per metric ton of initial uranium (GWd/t). Measurements were performed mainly using inductively coupled plasma mass spectrometry after elemental separation via highmore » performance liquid chromatography. High precision measurements were achieved using isotope dilution techniques for many of the lanthanides, uranium, and plutonium isotopes. Measurements are reported for more than 50 different isotopes and 16 elements. One of the two TMI-1 fuel rods measured in this work had been measured previously by Argonne National Laboratory (ANL), and these data have been widely used to support code and nuclear data validation. Recently, ORNL provided an important opportunity to independently cross check results against previous measurements performed at ANL. The measured nuclide concentrations are used to validate burnup calculations using the SCALE nuclear systems modeling and simulation code suite. These results show that the new measurements provide reliable benchmark data for computer code validation.« less

  8. EIS-0015: U.S. Spent Fuel Policy

    Broader source: Energy.gov [DOE]

    Subsumed DOE/EIS-0040 and DOE/EIS-0041. The Savannah River Laboratory prepared this EIS to analyze the impacts of implementing or not implementing the policy for interim storage of spent power reactor fuel. This Final EIS is a compilation of three Draft EISs and one Supplemental Draft EIS: DOE/EIS-0015-D, Storage of U.S. Spent Power Reactor Fuel; DOE/EIS-0015-DS, Storage of U.S. Spent Power Reactor Fuel - Supplement; DOE/EIS-0040-D, Storage of Foreign Spent Power Reactor Fuel; and DOE/EIS-0041-D, Charge for Spent Fuel Storage.

  9. Decomposition of chlorinated ethylenes and ethanes in an electron beam generated plasma reactor

    SciTech Connect (OSTI)

    Vitale, S.A.

    1996-02-01

    An electron beam generated plasma reactor (EBGPR) is used to determine the plasma chemistry kinetics, energetics and decomposition pathways of six chlorinated ethylenes and ethanes: 1,1,1-trichloroethane, 1,1-dichloroethane, ethyl chloride, trichloroethylene, 1,1-dichloroethylene, and vinyl chloride. A traditional chemical kinetic and chemical engineering analysis of the data from the EBGPR is performed, and the following hypothesis was verified: The specific energy required for chlorinated VOC decomposition in the electron beam generated plasma reactor is determined by the electron attachment coefficient of the VOC and the susceptibility of the molecule to radical attack. The technology was demonstrated at the Hanford Reservation to remove VOCs from soils.

  10. Considerations Associated with Reactor Technology Selection for the Next Generation Nuclear Plant Project

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01

    At the inception of the Next Generation Nuclear Plant Project and during predecessor activities, alternative reactor technologies have been evaluated to determine the technology that best fulfills the functional and performance requirements of the targeted energy applications and market. Unlike the case of electric power generation where the reactor performance is primarily expressed in terms of economics, the targeted energy applications involve industrial applications that have specific needs in terms of acceptable heat transport fluids and the associated thermodynamic conditions. Hence, to be of interest to these industrial energy applications, the alternative reactor technologies are weighed in terms of the reactor coolant/heat transport fluid, achievable reactor outlet temperature, and practicality of operations to achieve the very high reliability demands associated with the petrochemical, petroleum, metals and related industries. These evaluations have concluded that the high temperature gas-cooled reactor (HTGR) can uniquely provide the required ranges of energy needs for these target applications, do so with promising economics, and can be commercialized with reasonable development risk in the time frames of current industry interest – i.e., within the next 10-15 years.

  11. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    SciTech Connect (OSTI)

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  12. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    SciTech Connect (OSTI)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for both the PBMR and prismatic design. The main focus of this report is the RPV for both design concepts with emphasis on material selection.

  13. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The first experiment was inserted in the ATR in August 2009 and started its irradiation in September 2009. It is anticipated to complete its irradiation in early calendar 2011. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and the irradiation experience to date.

  14. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect (OSTI)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  15. Spent fuel storage alternatives

    SciTech Connect (OSTI)

    O'Connell, R.H.; Bowidowicz, M.A.

    1983-01-01

    This paper compares a small onsite wet storage pool to a dry cask storage facility in order to determine what type of spent fuel storage alternatives would best serve the utilities in consideration of the Nuclear Waste Policy Act of 1982. The Act allows the DOE to provide a total of 1900 metric tons (MT) of additional spent fuel storage capacity to utilities that cannot reasonably provide such capacity for themselves. Topics considered include the implementation of the Act (DOE away-from reactor storage), the Act's impact on storage needs, and an economic evaluation. The Waste Act mandates schedules for the determination of several sites, the licensing and construction of a high-level waste repository, and the study of a monitored retrievable storage facility. It is determined that a small wet pool storage facility offers a conservative and cost-effective approach for many stations, in comparison to dry cask storage.

  16. Report to Congress on Plan for Interim Storage of Spent Nuclear...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Nuclear Fuel from Decommissioned Reactors Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors PDF icon Report to Congress on Plan ...

  17. Active Interrogation for Spent Fuel

    SciTech Connect (OSTI)

    Swinhoe, Martyn Thomas; Dougan, Arden

    2015-11-05

    The DDA instrument for nuclear safeguards is a fast, non-destructive assay, active neutron interrogation technique using an external 14 MeV DT neutron generator for characterization and verification of spent nuclear fuel assemblies.

  18. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect (OSTI)

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  19. Development and Transient Analysis of a Helical-coil Steam Generator for High Temperature Reactors

    SciTech Connect (OSTI)

    Nathan V. Hoffer; Nolan A. Anderson; Piyush Sabharwall

    2011-08-01

    A high temperature gas-cooled reactor (HTGR) is under development by the Next Generation Nuclear Plant (NGNP) Project at the Idaho National Laboratory (INL). Its design emphasizes electrical power production which may potentially be coupled with process heat for hydrogen production and other industrial applications. NGNP is considering a helical-coil steam generator for the primary heat transport loop heat exchanger based on its increased heat transfer and compactness when compared to other steam generators. The safety and reliability of the helical-coil steam generator is currently under evaluation as part of the development of NGNP. Transients, such as loss of coolant accidents (LOCA), are of interest in evaluating the safety of steam generators. In this study, a complete steam generator inlet pipe break (double ended pipe break) LOCA was simulated by an exponential loss of primary side pressure. For this analysis, a model of the helical-coil steam generator was developed using RELAP5-3D, an INL inhouse systems analysis code. The steam generator model behaved normally during the transient simulating the complete steam generator inlet pipe break LOCA. Further analysis is required to comprehensively evaluate the safety and reliability of the helical-coil steam generator design in the NGNP setting.

  20. Actinide removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C. (Pleasanton, CA); von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

    2002-01-01

    A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

  1. Metals removal from spent salts

    DOE Patents [OSTI]

    Hsu, Peter C. (Pleasanton, CA); Von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Brummond, William A. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

    2002-01-01

    A method and apparatus for removing metal contaminants from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 1.0 ppm of contaminants.

  2. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    SciTech Connect (OSTI)

    Gougar, Hans David

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  3. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOE Patents [OSTI]

    Lasche, G.P.

    1983-09-29

    The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

  4. Steam generator materials performance in high temperature gas-cooled reactors

    SciTech Connect (OSTI)

    Chafey, J.E.; Roberts, D.I.

    1980-11-01

    This paper reviews the materials technology aspects of steam generators for HTGRs which feature a graphite-moderated, uranium-thorium, all-ceramic core and utilizes high-pressure helium as the primary coolant. The steam generators are exposed to gas-side temperatures approaching 760/sup 0/C and produce superheated steam at 538/sup 0/C and 16.5 MPa (2400 psi). The prototype Peach Bottom I 40-MW(e) HTGR was operated for 1349 EFPD over 7 years. Examination after decommissioning of the U-tube steam generators and other components showed the steam generators to be in very satisfactory condition. The 330-MW(e) Fort St. Vrain HTGR, now in the final stages of startup, has achieved 70% power and generated more than 1.5 x 10/sup 6/ MWh of electricity. The steam generators in this reactor are once-through units of helical configuration, requiring a number of new materials factors including creep-fatigue and water chemistry control. Current designs of larger HTGRs also feature steam generators of helical once-through design. Materials issues that are important in these designs include detailed consideration of time-dependent behavior of both base metals and welds, as required by current American Society of Mechanical Engineers (ASME) Code rules, evaluation of bimetallic weld behavior, evaluation of the properties of large forgings, etc.

  5. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    DOE Patents [OSTI]

    Fisch, Nathaniel J. (Cambridge, MA)

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma.

  6. System and method for generating steady state confining current for a toroidal plasma fusion reactor

    DOE Patents [OSTI]

    Bers, Abraham (Arlington, MA)

    1981-01-01

    A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma.

  7. Report on interim storage of spent nuclear fuel

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  8. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

  9. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  10. Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Decommissioned Reactors | Department of Energy Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors PDF icon Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors More Documents & Publications Information Request, "THE REPORT TO THE PRESIDENT AND THE CONGRESS BY THE SECRETARY OF ENERGY ON THE NEED

  11. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING DEACTIVATION AND DECOMMISSIONING OF REACTOR VESSELS AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Wiersma, B.; Serrato, M.; Langton, C.

    2010-11-10

    The R- and P-reactor vessels at the Savannah River Site (SRS) are being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of physically isolating and stabilizing the reactor vessel by filling it with a grout material. The reactor vessels contain aluminum alloy materials, which pose a concern in that aluminum corrodes rapidly when it comes in contact with the alkaline grout. A product of the corrosion reaction is hydrogen gas and therefore potential flammability issues were assessed. A model was developed to calculate the hydrogen generation rate as the reactor is being filled with the grout material. Three options existed for the type of grout material for D&D of the reactor vessels. The grout formulation options included ceramicrete (pH 6-8), a calcium aluminate sulfate (CAS) based cement (pH 10), or Portland cement grout (pH 12.4). Corrosion data for aluminum in concrete were utilized as input for the model. The calculations considered such factors as the surface area of the aluminum components, the open cross-sectional area of the reactor vessel, the rate at which the grout is added to the reactor vessel, and temperature. Given the hydrogen generation rate, the hydrogen concentration in the vapor space of the reactor vessel above the grout was calculated. This concentration was compared to the lower flammability limit for hydrogen. The assessment concluded that either ceramicrete or the CAS grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters did not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. Therefore, it was recommended that this grout not be utilized for this task. On the other hand, the R-reactor vessel contained significantly less aluminum surface area that the P-reactor vessel based on current facility process knowledge, surface observations, and drawings. Therefore, a Portland cement grout may be considered for grouting operations as well as the other grout formulations.

  12. REACTORS

    DOE Patents [OSTI]

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  13. Recovery of tritium dissolved in sodium at the steam generator of fast breeder reactor

    SciTech Connect (OSTI)

    Oya, Y.; Oda, T.; Tanaka, S.; Okuno, K.

    2008-07-15

    The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D{sub 2} gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 1/200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment. (authors)

  14. Regulatory Concerns on the In-Containment Water Storage System of the Korean Next Generation Reactor

    SciTech Connect (OSTI)

    Ahn, Hyung-Joon; Lee, Jae-Hun; Bang, Young-Seok; Kim, Hho-Jung

    2002-07-15

    The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety and Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified.

  15. Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors

    SciTech Connect (OSTI)

    Iouri I. Balachov; Takao Kobayashi; Francis Tanzella; Indira Jayaweera; Palitha Jayaweera; Petri Kinnunen; Martin Bojinov; Timo Saario

    2004-11-17

    This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.

  16. Ozone generation by negative direct current corona discharges in dry air fed coaxial wire-cylinder reactors

    SciTech Connect (OSTI)

    Yehia, Ashraf; Mizuno, Akira

    2013-05-14

    An analytical study was made in this paper for calculating the ozone generation by negative dc corona discharges. The corona discharges were formed in a coaxial wire-cylinder reactor. The reactor was fed by dry air flowing with constant rates at atmospheric pressure and room temperature, and stressed by a negative dc voltage. The current-voltage characteristics of the negative dc corona discharges formed inside the reactor were measured in parallel with concentration of the generated ozone under different operating conditions. An empirical equation was derived from the experimental results for calculating the ozone concentration generated inside the reactor. The results, that have been recalculated by using the derived equation, have agreed with the experimental results over the whole range of the investigated parameters, except in the saturation range for the ozone concentration. Therefore, the derived equation represents a suitable criterion for expecting the ozone concentration generated by negative dc corona discharges in dry air fed coaxial wire-cylinder reactors under any operating conditions in range of the investigated parameters.

  17. Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

    SciTech Connect (OSTI)

    Vierow, Karen; Aldemir, Tunc

    2009-09-10

    The project entitled, “Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors”, was conducted as a DOE NERI project collaboration between Texas A&M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

  18. Waste management plan for Hanford spent nuclear fuel characterization activities

    SciTech Connect (OSTI)

    Chastain, S.A. [Westinghouse Hanford Co., Richland, WA (United States); Spinks, R.L. [Pacific Northwest Lab., Richland, WA (United States)

    1994-10-17

    A joint project was initiated between Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratory (PNL) to address critical issues associated with the Spent Nuclear Fuel (SNF) stored at the Hanford Site. Recently, particular attention has been given to remediation of the SNF stored in the K Basins. A waste management plan (WMP) acceptable to both parties is required prior to the movement of selected material to the PNL facilities for examination. N Reactor and Single Pass Reactor (SPR) fuel has been stored for an extended period of time in the N Reactor, PUREX, K-East, and K-West Basins. Characterization plans call for transport of fuel material form the K Basins to the 327 Building Postirradiation Testing Laboratory (PTL) in the 300 Area for examination. However, PNL received a directive stating that no examination work will be started in PNL hot cell laboratories without an approved disposal route for all waste generated related to the activity. Thus, as part of the Characterization Program Management Plan for Hanford Spent Nuclear Fuel, a waste management plan which will ensure that wastes generated as a result of characterization activities conducted at PNL will be accepted by WHC for disposition is required. This document contains the details of the waste handling plan that utilizes, to the greatest extent possible, established waste handling and disposal practices at Hanford between PNL and WHC. Standard practices are sufficient to provides for disposal of most of the waste materials, however, special consideration must be given to the remnants of spent nuclear fuel elements following examination. Fuel element remnants will be repackaged in an acceptable container such as the single element canister and returned to the K Basins for storage.

  19. Evaluation of the potential for fish passage through the N Reactor and the Hanford generating project discharges

    SciTech Connect (OSTI)

    Dauble, D.D.; Vail, L.W.; Neitzel, D.A.

    1987-09-01

    The potential for juvenile downstream-migrating salmonids to encounter both the Hanford Generating Project (HGP) and N Reactor discharges was evaluated. Three general scenarios were assessed for fish exposure: (1) HGP plume centerline passage followed by N Reator plum centerline passage, (2) HGP plume centerline passage including intersection with the N Reactor plume, and (3) noncenterline plume passage through the edge of first the HGP and then the N Reactor plume. It is highly unlikely that a fish would pass through both plume centerlines because of the location of the two discharges and because of river-mixing characteristics near the discharges. For the set of conditions that we evaluated, exposure to elevated temperatures would be of insufficient duration to result in mortalities to fish that might encounter both the HGP and N Reactor plumes.

  20. An overview of spent-fuel processing in the global nuclear-energy partnership

    SciTech Connect (OSTI)

    Laidler, James J.

    2008-07-01

    Spent nuclear fuel is being generated at a prodigious rate in the U.S. and in other countries with robust nuclear-power-generation infrastructures, and the annual rate of production is likely to triple by 2050. The U.S. is engaged in the development of commercial light-water-reactor spent- fuel-treatment processes that are intended to meet certain rigorous criteria for separations efficiency, waste management benefits, and economy of industrial-scale operations. Aqueous solvent-extraction processes are the technology of choice, and a variety of process options have been designed and tested for technical feasibility. In general, the processes involve substantial partitioning of the constituents of spent nuclear fuel, so that effective use can be made of the recovered unburned uranium and other fissile isotopes that can be recycled as fuel for contemporary or advanced reactors. Those constituents that are destined for disposal as waste are also separated in order that they can be placed into durable waste forms that are expressly tailored for a particular disposition pathway. The U.S. is also working with international partners as part of the Global Nuclear Energy Partnership (GNEP) to develop a consistent worldwide approach to the treatment of spent fuel and the disposition of wastes arising from such processing. (authors)

  1. Waste Stream Generated and Waste Disposal Plans for Molten Salt Reactor Experiment at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Haghighi, M. H.; Szozda, R. M.; Jugan, M. R.

    2002-02-26

    The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR), south of the Oak Ridge National Laboratory (ORNL) main plant across Haw Ridge in Melton Valley. The MSRE was run by ORNL to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503 (Figure 1). The reactor was operated from June 1965 to December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed t o cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. As a result of the S&M program, it was discovered in 1994 that gaseous uranium (233U/232U) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 was generated when radiolysis of the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine.Some of the free fluorine combined with uranium fluorides (UF4) in the salt to form UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE.

  2. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R- AND P-REACTOR VESSELS

    SciTech Connect (OSTI)

    Wiersma, B.

    2009-12-29

    The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a significant margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel contains significantly less aluminum and thus a Portland cement grout may be considered as well. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation in the R-reactor vessel is very low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained at a temperature of 80 C, the risk will again be very low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. For P-reactor, grout temperatures less than 100 C should provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. For R-reactor, grout temperatures less than 70 C or 80 C will provide an adequate safety margin for the Portland cement. The other grout formulations are also viable options for R-reactor. (2) Minimize the grout fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. For P-reactor, fill rates that are less than 2 inches/min for the ceramicrete and the silica fume grouts will reduce the chance of significant hydrogen accumulation. For R-reactor, fill rates less than 1 inch/min will again minimize the risk of hydrogen accumulation. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates in the P-reactor vessel, however, are low for the pH 8 and pH 10.4 grout, (i.e., less than 0.32 ft{sup 3}/min). If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

  3. Technical bases for interim storage of spent nuclear fuel

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.

    1981-06-01

    The experience base for water storage of spent nuclear fuel has evolved since 1943. The technology base includes licensing documentation, standards, technology studies, pool operator experience, and documentation from public hearings. That base reflects a technology which is largely successful and mundane. It projects probable satisfactory water storage of spent water reactor fuel for several decades. Interim dry storage of spent water reactor fuel is not yet licensed in the US, but a data base and documentation have developed. There do not appear to be technological barriers to interim dry storage, based on demonstrations with irradiated fuel. Water storage will continue to be a part of spent fuel management at reactors. Whether dry storage becomes a prominent interim fuel management option depends on licensing and economic considerations. National policies will strongly influence how long the spent fuel remains in interim storage and what its final disposition will be.

  4. Accelerator-driven transmutation of spent fuel elements

    DOE Patents [OSTI]

    Venneri, Francesco; Williamson, Mark A.; Li, Ning

    2002-01-01

    An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

  5. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  6. Steam Generator Component Model in a Combined Cycle of Power Conversion Unit for Very High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Oh, Chang H; Han, James; Barner, Robert; Sherman, Steven R

    2007-06-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP), Very High Temperature Gas-Cooled Reactor (VHTR) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. A combined cycle is considered as one of the power conversion units to be coupled to the very high-temperature gas-cooled reactor (VHTR). The combined cycle configuration consists of a Brayton top cycle coupled to a Rankine bottoming cycle by means of a steam generator. A detailed sizing and pressure drop model of a steam generator is not available in the HYSYS processes code. Therefore a four region model was developed for implementation into HYSYS. The focus of this study was the validation of a HYSYS steam generator model of two phase flow correlations. The correlations calculated the size and heat exchange of the steam generator. To assess the model, those calculations were input into a RELAP5 model and its results were compared with HYSYS results. The comparison showed many differences in parameters such as the heat transfer coefficients and revealed the different methods used by the codes. Despite differences in approach, the overall results of heat transfer were in good agreement.

  7. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750–800°C Reactor Outlet Temperature

    SciTech Connect (OSTI)

    John Collins

    2009-08-01

    This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750–800°C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

  8. FORIG: a computer code for calculating radionuclide generation and depletion in fusion and fission reactors. User's manual

    SciTech Connect (OSTI)

    Blink, J.A.

    1985-03-01

    In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs.

  9. Spent Fuel Background Report Volume I

    SciTech Connect (OSTI)

    Abbott, D.

    1994-03-01

    This report is an overview of current spent nuclear fuel management in the DOE complex. Sources of information include published literature, internal DOE documents, interviews with site personnel, and information provided by individual sites. Much of the specific information on facilities and fuels was provided by the DOE sites in response to the questionnaire for data for spent fuels and facilities data bases. This information is as accurate as is currently available, but is subject to revision pending results of further data calls. Spent fuel is broadly classified into three categories: (a) production fuels, (b) special fuels, and (c) naval fuels. Production fuels, comprising about 80% of the total inventory, are those used at Hanford and Savannah River to produce nuclear materials for defense. Special fuels are those used in a wide variety of research, development, and testing activities. Special fuels include fuel from DOE and commercial reactors used in research activities at DOE sites. Naval fuels are those developed and used for nuclear-powered naval vessels and for related research and development. Given the recent DOE decision to curtail reprocessing, the topic of main concern in the management of spent fuel is its storage. Of the DOE sites that have spent nuclear fuel, the vast majority is located at three sites-Hanford, INEL, and Savannah River. Other sites with spent fuel include Oak Ridge, West Valley, Brookhaven, Argonne, Los Alamos, and Sandia. B&W NESI Lynchburg Technology Center and General Atomics are commercial facilities with DOE fuel. DOE may also receive fuel from foreign research reactors, university reactors, and other commercial and government research reactors. Most DOE spent fuel is stored in water-filled pools at the reactor facilities. Currently an engineering study is being performed to determine the feasibility of using dry storage for DOE-owned spent fuel currently stored at various facilities. Delays in opening the deep geologic repository and the decision to phase out reprocessing of production fuels are extending the need for interim storage. The report describes the basic storage conditions and the general SNF inventory at individual DOE facilities.

  10. Generation IV International Forum Signs Agreement to Collaborate on Sodium Cooled Fast Reactors

    Broader source: Energy.gov [DOE]

    FUKUI , JAPAN - The Department of Energy today announced that the United States signed a sodium-cooled fast reactor systems arrangement with France and Japan, providing the framework for...

  11. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    SciTech Connect (OSTI)

    Tan, Lizhen; Yang, Ying; Sridharan, K.

    2015-12-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  12. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    SciTech Connect (OSTI)

    Lasche, G.P.

    1988-04-05

    A method for recovering energy in an inertial confinement fusion reactor having a reactor chamber and a sphere forming means positioned above an opening in the reactor chamber is described, comprising: embedding a fusion target fuel capsule having a predetermined yield in the center of a hollow solid lithium tube and subsequently embedding the hollow solid lithium tube in a liquid lithium medium; using the sphere forming means for forming the liquid lithium into a spherical shaped liquid lithium mass having a diameter smaller than the length of the hollow solid lithium tube with the hollow solid lithium tube being positioned along a diameter of the spherical shaped mass, providing the spherical shaped liquid lithium mass with the fusion fuel target capsule and hollow solid lithium tube therein as a freestanding liquid lithium shaped spherical shaped mass without any external means for maintaining the spherical shape by dropping the liquid lithium spherical shaped mass from the sphere forming means into the reactor chamber; producing a magnetic field in the reactor chamber; imploding the target capsule in the reactor chamber to produce fusion energy; absorbing fusion energy in the liquid lithium spherical shaped mass to convert substantially all the fusion energy to shock induced kinetic energy of the liquid lithium spherical shaped mass which expands the liquid lithium spherical shaped mass; and compressing the magnetic field by expansion of the liquid lithium spherical shaped mass and recovering useful energy.

  13. Evaluation of the Strategic Value of Fully Burnt PBMR Spent Fuel - A Report to ISPO in Response to IAEA Letter Request (2004-08-30)

    SciTech Connect (OSTI)

    A. M. Ougouag; H. D. Gougar; T. A. Todd

    2006-05-01

    The IAEA needs to determine the value of imposing safeguards on the spent fuel storage at the Pebble Bed Modular Reactor (PBMR) planned for construction in the Republic of South Africa. The PBMR will use hundreds of thousands of fuel elements in the shape of small spheres (6 cm in diameter). The PBMR plant design calls for the storage on site of all the spent fuel generated during the whole life of the reactor, expected to span 40 years. The spent fuel storage system is designed (or to be designed) for a functional life of 80 years. If it is determined that the spent fuel contains materials of interest to a would-be proliferant, then safeguards would have to be imposed and maintained until the spent fuel elements are processed into a form and composition that no longer requires safeguards. The problem addressed in this report is the determination of the strategic value of the spent fuel to such a would-be proliferant.

  14. Historical overview of domestic spent fuel shipments: Update

    SciTech Connect (OSTI)

    Not Available

    1991-07-01

    This report presents available historic data on most commercial and research reactor spent fuel shipments in the United States from 1964 through 1989. Data include sources of the spent fuel shipped, types of shipping casks used, number of fuel assemblies shipped, and number of shipments made. This report also addresses the shipment of spent research reactor fuel. These shipments have not been documented as well as commercial power reactor spent fuel shipment activity. Available data indicate that the greatest number of research reactor fuel shipments occurred in 1986. The largest campaigns in 1986 were from the Brookhaven National Laboratory, Brooklyn, New York, to the Idaho Chemical Processing Plant (ICPP) and from the Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) in Tennessee and the Rockwell International Reactor in California to the Savannah River Plant near Aiken, South Carolina. For all years addressed in this report, DOE facilities in Idaho Falls and Savannah River were the major recipients of research reactor spent fuel. In 1989, 10 shipments were received at the Idaho facilities. These originated from universities in California, Michigan, and Missouri. 9 refs., 12 figs., 7 tabs.

  15. Characterization plan for Hanford spent nuclear fuel

    SciTech Connect (OSTI)

    Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

    1994-12-01

    Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF.

  16. Investigation of particulate corrosion product transients in the primary coolant of the Winfrith steam generating heavy water reactor

    SciTech Connect (OSTI)

    Means, F.A.; Rodliffe, R.S.; Harding, K.

    1980-03-01

    Equipment for on-line counting and sizing of particles has been used to sample coolant from the primary circuit of a water reactor (the Winfrith steam generating heavy water reactor). The particle size distribution is compared with a determination by electron microscopic examination of a filter sample and is shown to be in good agreement. The technique allows transients in coolant-borne particle concentrations to be sufficiently resolved for analysis in terms of postulated particle deposition and resuspension behavior. The deposition behavior is found to be describable by a first-order rate process with rate constants smaller than those that would be predicted from mass transfer considerations. It is concluded that deposition cannot be limited by mass transfer alone.

  17. Report on interim storage of spent nuclear fuel. Midwestern high-level radioactive waste transportation project

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  18. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

    SciTech Connect (OSTI)

    Wiersma, B.

    2010-05-24

    The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D and D). D and D activities consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS and T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D and D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement groupt (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel cotnains significantly less aluminum based on current facility process knowledge, surface observations, and drawings. Therefore, a Portland cement grout may be considered for grouting operations as well as the other grout formulations. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation during fill operations in the R-reactor vessel is low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained at a temperature of 80 C, the risk is again low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. For P-reactor, grout temperatures less than 100 C should provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. For R-reactor, grout temperatures less than 70 C or 80 C will provide an adequate safety margin for the Portland cement. The other grout formulations are also viable options for R-reactor. (2) Minimize the grout fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. For P-reactor, fill rates that are less than 2 inches/min for the ceramicrete and the silica fume grouts will reduce the chance of significant hydrogen accumulation. For R-reactor, fill rates less than 1 inch/min will again minimize the risk of hydrogen accumulation. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates in the P-reactor vessel, however, are low for the pH 8 and pH 10.4 grout, (i.e., less than 0.97 ft{sup 3}/min). If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

  19. ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

    SciTech Connect (OSTI)

    Wiersma, B.

    2009-10-29

    The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D&D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or portland cement grout (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Conservative calculations estimate that either ceramicrete or the silica fume grout may be used to safely grout the R- and P- reactor vessels. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. Grout temperatures less than 100 C should however, still provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. (2) Minimize the fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. Fill rates that are less than 2 inches/min will reduce the chance of significant hydrogen build-up. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates however, are low for the pH 8 and pH 10.4 grout, i.e., less than 0.32 ft{sup 3}/min. Portland cement grout, on the other hand, for the same range of process parameters does not provide a significant margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations. It is recommended that this grout not be utilized for this task. If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

  20. Spent Nuclear Fuel Alternative Technology Decision Analysis

    SciTech Connect (OSTI)

    Shedrow, C.B.

    1999-11-29

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

  1. Characterization of radioactive aerosols generated during the demolition of the concrete bioshield of a reactor undergoing decommissioning

    SciTech Connect (OSTI)

    Landolt, R.R.

    1996-12-31

    Particle size distribution studies were performed on aerosols generated during the cutting of the core support floor of the Ft. St. Vrain reactor which is undergoing decommissioning. The activity median aerodynamic diameter (AMAD) for Co-60 in the aerosol was found to be considerably larger than the currently used 1-micron default value and in one case larger than the newly recommended 5-micron reference value for occupational exposure. The reverse was true for Cs-137 where the AMAD was found to be in the sub-micron range possibly caused by high solubility of the radiocesium into the water spray used in the cutting operation.

  2. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOE Patents [OSTI]

    Lasche, George P. (Arlington, VA)

    1988-01-01

    A high-power-density laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems.

  3. Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

    DOE Patents [OSTI]

    Lasche, G.P.

    1987-02-20

    A high-power-density-laser or charged-particle-beam fusion reactor system maximizes the directed kinetic energy imparted to a large mass of liquid lithium by a centrally located fusion target. A fusion target is embedded in a large mass of lithium, of sufficient radius to act as a tritium breeding blanket, and provided with ports for the access of beam energy to implode the target. The directed kinetic energy is converted directly to electricity with high efficiency by work done against a pulsed magnetic field applied exterior to the lithium. Because the system maximizes the blanket thickness per unit volume of lithium, neutron-induced radioactivities in the reaction chamber wall are several orders of magnitude less than is typical of other fusion reactor systems. 25 figs.

  4. Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

    SciTech Connect (OSTI)

    Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

    2012-07-01

    700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

  5. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    SciTech Connect (OSTI)

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  6. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods

    SciTech Connect (OSTI)

    Johnson, G.L.

    1988-09-01

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store lightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degree}C and whether the cladding of the stored spent fuel ever exceeds 350{degree}C. Limiting the borehole to temperatures of 97{degree}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degree}C cladding limit minimizes the possibility of creep-related failure in the spent fuel rod cladding. For a series of packages stored in a 8 x 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degree}C for the full 1000-yr analysis period.

  7. TEPP- Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    This scenario provides the planning instructions, guidance, and evaluation forms necessary to conduct an exercise involving a highway shipment of spent nuclear fuel.  This exercise manual is one in...

  8. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    SciTech Connect (OSTI)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-06

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  9. Arrival condition of spent fuel after storage, handling, and transportation

    SciTech Connect (OSTI)

    Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

    1982-11-01

    This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables.

  10. Tritium permeation characterization of materials for fusion and generation IV very high temperature reactors

    SciTech Connect (OSTI)

    Thomson, S.; Pilatzke, K.; McCrimmon, K.; Castillo, I.; Suppiah, S.

    2015-03-15

    The objective of this work is to establish the tritium-permeation properties of structural alloys considered for Fusion systems and very high temperature reactors (VHTR). A description of the work performed to set up an apparatus to measure permeation rates of hydrogen and tritium in 304L stainless steel is presented. Following successful commissioning with hydrogen, the test apparatus was commissioned with tritium. Commissioning tests with tritium suggest the need for a reduction step that is capable of removing the oxide layer from the test sample surfaces before accurate tritium-permeation data can be obtained. Work is also on-going to clearly establish the temperature profile of the sample to correctly estimate the tritium-permeability data.

  11. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA-Winfrith

    SciTech Connect (OSTI)

    Miller, Keith; Cornell, Rowland; Parkinson, Steve; McIntyre, Kevin; Staples, Andy

    2007-07-01

    Available in abstract form only. Full text of publication follows: The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and all remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height ISO containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. (authors)

  12. Decommissioning of the secondary containment of the steam generating heavy water reactor at UKAEA Winfrith

    SciTech Connect (OSTI)

    Miller, K.D.; Cornell, R.M.; Parkinson, S.J.; McIntyre, K.; Staples, A.

    2007-07-01

    The Winfrith SGHWR was a prototype nuclear power plant operated for 23 years by the United Kingdom Atomic Energy Authority (UKAEA) until 1990 when it was shut down permanently. The current Stage 1 decommissioning contract is part of a multi-stage strategy. It involves the removal of all the ancillary plant and equipment in the secondary containment and non-containment areas ahead of a series of contracts for the decommissioning of the primary containment, the reactor core and demolition of the building and ail remaining facilities. As an outcome of a competitive tending process, the Stage 1 decommissioning contract was awarded to NUKEM with operations commencing in April 2005. The decommissioning processes involved with these plant items will be described with some emphasis of the establishment of multiple work-fronts for the production, recovery, treatment and disposal of mainly tritium-contaminated waste arising from its contact with the direct cycle reactor coolant. The means of size reduction of a variety of large, heavy and complex items of plant made from a range of materials will also be described with some emphasis on the control of fumes during hot cutting operations and establishing effective containments within a larger secondary containment structure. Disposal of these wastes in a timely and cost-effective manner is a major challenge facing the decommissioning team and has required the development of a highly efficient means of packing the resultant materials into mainly one-third height IS0 containers for disposal as LLW. Details of the quantities of LLW and exempt wastes handled during this process will be given with a commentary about the difficulty in segregating these two waste streams efficiently. The paper sets out to demonstrate the considerable progress that has been made with these challenging decommissioning operations at the SGHWR plant and to highlight some of the techniques and processes that have contributed to the overall success of the process. The overall management and control of safety during ail aspects of this challenging contract are major features of the paper, greatly assisted by the adoption from the outset of a non-adversarial team working approach between client and contractor. This has greatly assisted in developing safe and cost-effective solutions to a number of problems that have arisen during the programme, demonstrating the worth of adopting this co-operative approach for mutual benefit. (authors)

  13. validation and Enhancement of Computational Fluid Dynamics and Heat Transfer Predictive Capabilities for Generation IV Reactor Systems

    SciTech Connect (OSTI)

    Robert E. Spall; Barton Smith; Thomas Hauser

    2008-12-08

    Nationwide, the demand for electricity due to population and industrial growth is on the rise. However, climate change and air quality issues raise serious questions about the wisdom of addressing these shortages through the construction of additional fossil fueled power plants. In 1997, the President's Committee of Advisors on Science and Technology Energy Research and Development Panel determined that restoring a viable nuclear energy option was essential and that the DOE should implement a R&D effort to address principal obstacles to achieving this option. This work has addressed the need for improved thermal/fluid analysis capabilities, through the use of computational fluid dynamics, which are necessary to support the design of generation IV gas-cooled and supercritical water reactors.

  14. Decomposition of 1,1-Dichloroethane and 1,1-Dichloroethene in an electron beam generated plasma reactor

    SciTech Connect (OSTI)

    Vitale, S.A.; Hadidi, K.; Cohn, D.R.; Bromberg, L.

    1997-03-01

    An electron beam generated plasma reactor is used to decompose low concentrations (100{endash}3000 ppm) of 1,1-dichloroethane and 1,1-dichloroethene in atmospheric pressure air streams. The energy requirements for 90{percent} and 99{percent} decomposition of each compound are reported as a function of inlet concentration. Dichloroethene decomposition is enhanced by a chlorine radical propagated chain reaction. The chain length of the dichloroethene reaction is estimated to increase with dichloroethene concentration from 10 at 100 ppm initial dichloroethene concentration to 30 at 3000 ppm. Both the dichloroethane and dichloroethene reactions seem to be inhibited by electron scavenging decomposition products. A simple analytic expression is proposed for fitting decomposition data where inhibition effects are important and simple first order kinetics are not observed. {copyright} {ital 1997 American Institute of Physics.}

  15. Surveillance of LWR spent fuel in wet storage. Final report, October 1984

    SciTech Connect (OSTI)

    Bailey, W.J.; Johnson, A.B. Jr.

    1984-10-01

    Battelle, Pacific Northwest Laboratories established a surveillance program for EPRI that documents the integrity of spent light-water reactor fuel and structural materials (spent fuel storage pool liners, racks, piping, etc.) during wet storage. The program involves providing an update on the overall performance of spent fuel in wet storage, monitoring Licensee Event Reports (LERs) for pertinent significant occurrences, identifying lead spent fuel assemblies that are of particular interest to EPRI, monitoring developments in fuel design and performance and assessing their influence on spent fuel storage characteristics, and identifying specific actions or programs that may be needed to maintain the viability of wet storage of spent fuel for extended periods. Experience to date indicates that wet storage is a well-developed technology with no associated major technological problems. Spent fuel storage pools are operated without substantial risk to the public or the plant personnel. A list of lead spent fuel assemblies is presented. Pertinent occurrences from LERs are listed. Very few fuel assemblies have suffered major mechanical damage as a result of handling operations at spent fuel storage pools. Experience to date with handling operations at spent fuel storage pools indicates that failed fuel rods and inadvertent fracturing of fuel rods can be accommodated. Minor problems have occurred with spent fuel storage pool components such as liners, racks, and piping. Surveillance continues to be needed on: (1) possible effects on handling and storage of spent fuel from extended burnup, hydrogen injection at boiling water reactors, and rod consolidation operations; (2) extended pool exposure of neutron-absorbing materials; (3) cracking of spent fuel storage pool piping at pressurized water reactors; and (4) control of impurities in spent fuel pool waters. 120 references, 13 figures, 10 tables.

  16. Temperature for Spent Fuel Dry Storage

    Energy Science and Technology Software Center (OSTI)

    1992-07-13

    DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) calculates allowable initial temperatures for dry storage of light-water-reactor spent fuel and the cumulative damage fraction of Zircaloy cladding for specified initial storage temperature and stress and cooling histories. It is made available to ensure compliance with NUREG 10CFR Part 72, Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI). Although the program''s principal purpose is to calculate estimatesmore » of allowable temperature limits, estimates for creep strain, annealing fraction, and life fraction as a function of storage time are also provided. Equations for the temperature of spent fuel in inert and nitrogen gas storage are included explicitly in the code; in addition, an option is included for a user-specified cooling history in tabular form, and tables of the temperature and stress dependencies of creep-strain rate and creep-rupture time for Zircaloy at constant temperature and constant stress or constant ratio of stress/modulus can be created. DATING includes the GEAR package for the numerical solution of the rate equations and DPLOT for plotting the time-dependence of the calculated cumulative damage-fraction, creep strain, radiation damage recovery, and temperature decay.« less

  17. Temperature for Spent Fuel Dry Storage

    Energy Science and Technology Software Center (OSTI)

    1992-07-13

    DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) calculates allowable initial temperatures for dry storage of light-water-reactor spent fuel and the cumulative damage fraction of Zircaloy cladding for specified initial storage temperature and stress and cooling histories. It is made available to ensure compliance with NUREG 10CFR Part 72, Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI). Although the program''s principal purpose is to calculate estimatesmore »of allowable temperature limits, estimates for creep strain, annealing fraction, and life fraction as a function of storage time are also provided. Equations for the temperature of spent fuel in inert and nitrogen gas storage are included explicitly in the code; in addition, an option is included for a user-specified cooling history in tabular form, and tables of the temperature and stress dependencies of creep-strain rate and creep-rupture time for Zircaloy at constant temperature and constant stress or constant ratio of stress/modulus can be created. DATING includes the GEAR package for the numerical solution of the rate equations and DPLOT for plotting the time-dependence of the calculated cumulative damage-fraction, creep strain, radiation damage recovery, and temperature decay.« less

  18. Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

    SciTech Connect (OSTI)

    Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

    1988-12-01

    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO/sub 2/ oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO/sub 2/ pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs.

  19. Further Evaluation of the Neutron Resonance Transmission Analysis (NRTA) Technique for Assaying Plutonium in Spent Fuel

    SciTech Connect (OSTI)

    J. W. Sterbentz; D. L. Chichester

    2011-09-01

    This is an end-of-year report (Fiscal Year (FY) 2011) for the second year of effort on a project funded by the National Nuclear Security Administration's Office of Nuclear Safeguards (NA-241). The goal of this project is to investigate the feasibility of using Neutron Resonance Transmission Analysis (NRTA) to assay plutonium in commercial light-water-reactor spent fuel. This project is part of a larger research effort within the Next-Generation Safeguards Initiative (NGSI) to evaluate methods for assaying plutonium in spent fuel, the Plutonium Assay Challenge. The second-year goals for this project included: (1) assessing the neutron source strength needed for the NRTA technique, (2) estimating count times, (3) assessing the effect of temperature on the transmitted signal, (4) estimating plutonium content in a spent fuel assembly, (5) providing a preliminary assessment of the neutron detectors, and (6) documenting this work in an end of the year report (this report). Research teams at Los Alamos National Laboratory (LANL), Lawrence Berkeley National Laboratory (LBNL), Pacific Northwest National Laboratory (PNNL), and at several universities are also working to investigate plutonium assay methods for spent-fuel safeguards. While the NRTA technique is well proven in the scientific literature for assaying individual spent fuel pins, it is a newcomer to the current NGSI efforts studying Pu assay method techniques having just started in March 2010; several analytical techniques have been under investigation within this program for two to three years or more. This report summarizes work performed over a nine month period from January-September 2011 and is to be considered a follow-on or add-on report to our previous published summary report from December 2010 (INL/EXT-10-20620).

  20. Thermal analysis for fuel handling system for sodium cooled reactor considering minor actinide-bearing metal fuel.

    SciTech Connect (OSTI)

    Chikazawa, Y.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues of handling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonably-sized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide-bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled-reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the sodium coolant. The cladding temperature requirement is maintained below the creep temperature limit to avoid any damage before core installation. The thermal analysis shows that a helium gas-filled cask can accommodate ABR-1000 fresh minor actinide-bearing fuel with 700-W decay heat. The above analysis results revealed the overall requirement for minor actinide-bearing metal fuel handling. The information is thought to be helpful in the design of the ABR-1000 and future sodium-cooled-reactor fuel-handling system.

  1. Impact analysis of spent fuel jacket assemblies

    SciTech Connect (OSTI)

    Aramayo, G.A.

    1994-06-01

    As part of the analyses performed in support of the reracking of the High Flux Isotope Reactor pool, it became necessary to prove the structural integrity of the spent fuel jacket assemblies subjected to gravity drop that result from postulated accidents associated with the handling of these assemblies while submerged in the pool. The spent fuel jacket assemblies are an integral part of the reracking project, and serve to house fuel assemblies. The structure integrity of the jacket assemblies from loads that result from impact from a height of 10 feet onto specified targets has been performed analytically using the computer program LS-DYNA3D. Nine attitudes of the assembly at the time of impact have been considered. Results of the analyses show that there is no failure of the assemblies as a result of the impact scenarios considered.

  2. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    SciTech Connect (OSTI)

    Rudisill, T; John Mickalonis, J

    2006-09-27

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO{sub 2}) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO{sub 2} layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH{sub 4}F)/ammonium nitrate (NH{sub 4}NO{sub 3}) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO{sub 2} layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH{sub 4}){sub 2}ZrF{sub 6}) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of contamination. The thermal decomposition of this material is also undesirable if the cladding hulls are melted for volume reduction or to produce waste forms. Handling and disposal of the corrosive off-gas stream and ZrO{sub 2}-containing dross must be addressed. The stability of Zr{sup 4+} in the NHF{sub 4}/NH{sub 4}NO{sub 3} solution is also a concern. Precipitation of ammonium zirconium fluorides upon cooling of the dissolving solution was observed in the feasibility experiments. Precipitation of the solids was attributed to the high fluoride to Zr ratios used in the experiments. The solubility of Zr{sup 4+} in NH{sub 4}F solutions decreases as the free fluoride concentration increases. The removal of the ZrO{sub 2} layer from Zircaloy-4 coupons with HF showed a strong dependence on both the concentration and temperature. Very rapid dissolution of the oxide layer and significant amounts of metal was observed in experiments using HF concentrations {ge} 2.5 M. Treatment of the coupons using HF concentrations {le} 1.0 M was very effective in removing the oxide layer. The most effective conditions resulted in dissolution rates which were less than approximately 2 mg/cm{sup 2}-min. With dissolution rates in this range, uniform removal of the oxide layer was obtained and a minimal amount of Zircaloy metal was dissolved. Future HF dissolution studies should focus on the decontamination of actual spent fuel cladding hulls to determine if the treated hulls meet criteria for disposal as a LLW.

  3. Re-evaluation of Spent Nuclear Fuel Assay Data for the Three...

    Office of Scientific and Technical Information (OSTI)

    Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation Gauld, Ian C. Oak Ridge National Lab. (ORNL), Oak Ridge,...

  4. Re-evaluation of Spent Nuclear Fuel Assay Data for the Three...

    Office of Scientific and Technical Information (OSTI)

    DOE PAGES Search Results Accepted Manuscript: Re-evaluation of Spent Nuclear Fuel Assay Data for the Three Mile Island Unit 1 Reactor and Application to Code Validation This...

  5. A Second Look at Neutron Resonance Transmission Analysis as a Spent Fuel NDA Technique

    SciTech Connect (OSTI)

    James W .Sterbentz; David L. Chichester

    2011-07-01

    Many different nondestructive analysis techniques are currently being investigated as a part of the United States Department of Energy's Next Generation Safeguards Initiative (NGSI) seeking methods to quantify plutonium in spent fuel. Neutron Resonance Transmission Analysis (NRTA) is one of these techniques. Having first been explored in the mid-1970s for the analysis of individual spent-fuel pins a second look, using advanced simulation and modeling methods, is now underway to investigate the suitability of the NRTA technique for assaying complete spent nuclear fuel assemblies. The technique is similar to neutron time-of-flight methods used for cross-section determinations but operates over only the narrow 0.1-20 eV range where strong, distinguishable resonances exist for both the plutonium (239, 240, 241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Initial modeling shows excellent agreement with previously published experimental data for measurements of individual spent-fuel pins where plutonium assays were demonstrated to have a precision of 2-4%. Within the simulation and modeling analyses of this project scoping studies have explored fourteen different aspects of the technique including the neutron source, drift tube configurations, and gross neutron transmission as well as the impacts of fuel burn up, cooling time, and fission-product interferences. These results show that NRTA may be a very capable experimental technique for spent-fuel assay measurements. The results suggest sufficient transmission strength and signal differentiability is possible for assays through up to 8 pins. For an 8-pin assay (looking at an assembly diagonally), 64% of the pins in a typical 17 ? 17 array of a pressurized water reactor fuel assembly can be part of a complete transmission assay measurement with high precision. Analysis of rows with up to 12 pins may also be feasible but with diminished precision. Preliminary data analysis of an NRTA simulation has demonstrated the simplicity of the technique.

  6. AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

    SciTech Connect (OSTI)

    Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all possible pool heat load conditions. - After 3 days, several different means are provided to continue spent fuel cooling using installed plant equipment as well as off-site equipment with built-in connections. Even for beyond design basis accidents with postulated pool damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 multiple spent fuel pool spray and fill systems provide additional lines of defense to prevent spent fuel damage. (authors)

  7. Supplement Analysis Â… Spent Nuclear Fuel and SRS H-Canyon Operations

    Energy Savers [EERE]

    DOE/EIS-0218-SA-07 SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM Highly Enriched Uranium Target Residue Material Transportation U.S. Department of Energy Washington, DC November 2015 DOE/EIS-0218-SA-07 SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM Highly Enriched Uranium Target Residue Material Transportation 1.0 INTRODUCTION The Department of Energy (DOE) has a continuing responsibility for safeguarding

  8. Pyrochemical Treatment of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    K. M. Goff; K. L. Howden; G. M. Teske; T. A. Johnson

    2005-10-01

    Over the last 10 years, pyrochemical treatment of spent nuclear fuel has progressed from demonstration activities to engineering-scale production operations. As part of the Advanced Fuel Cycle Initiative within the U.S. Department of Energy’s Office of Nuclear Energy, Science and Technology, pyrochemical treatment operations are being performed as part of the treatment of fuel from the Experimental Breeder Reactor II at the Idaho National Laboratory. Integral to these treatment operations are research and development activities that are focused on scaling further the technology, developing and implementing process improvements, qualifying the resulting high-level waste forms, and demonstrating the overall pyrochemical fuel cycle.

  9. Advanced Reactor Technologies | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Nuclear Reactor Technologies » Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies The Office of Advanced Reactor Technologies (ART) sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative

  10. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    SciTech Connect (OSTI)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  11. The united kingdom's changing requirements for spent fuel storage

    SciTech Connect (OSTI)

    Hodgson, Z.; Hambley, D.I.; Gregg, R.; Ross, D.N.

    2013-07-01

    The UK is adopting an open fuel cycle, and is necessarily moving to a regime of long term storage of spent fuel, followed by geological disposal once a geological disposal facility (GDF) is available. The earliest GDF receipt date for legacy spent fuel is assumed to be 2075. The UK is set to embark on a programme of new nuclear build to maintain a nuclear energy contribution of 16 GW. Additionally, the UK are considering a significant expansion of nuclear energy in order to meet carbon reduction targets and it is plausible to foresee a scenario where up to 75 GW from nuclear power production could be deployed in the UK by the mid 21. century. Such an expansion, could lead to spent fuel storage and its disposal being a dominant issue for the UK Government, the utilities and the public. If the UK were to transition a closed fuel cycle, then spent fuel storage should become less onerous depending on the timescales. The UK has demonstrated a preference for wet storage of spent fuel on an interim basis. The UK has adopted an approach of centralised storage, but a 16 GW new build programme and any significant expansion of this may push the UK towards distributed spent fuel storage at a number of reactors station sites across the UK.

  12. Numerical Estimation of the Spent Fuel Ratio

    SciTech Connect (OSTI)

    Lindgren, Eric R.; Durbin, Samuel; Wilke, Jason; Margraf, J.; Dunn, T. A.

    2016-01-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

  13. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect (OSTI)

    Not Available

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  14. Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

    SciTech Connect (OSTI)

    Vdovin, V.

    2014-02-12

    The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].

  15. Air Shipment of Spent Nuclear Fuel from Romania to Russia

    SciTech Connect (OSTI)

    Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

    2010-10-01

    Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

  16. MOX and MOX with 237Np/241Am Inert Fission Gas Generation Comparison in ATR

    SciTech Connect (OSTI)

    G. S. Chang; M. Robel; W. J. Carmack; D. J. Utterbeck

    2006-06-01

    The treatment of spent fuel produced in nuclear power generation is one of the most important issues to both the nuclear community and the general public. One of the viable options to long-term geological disposal of spent fuel is to extract plutonium, minor actinides (MA), and potentially long-lived fission products from the spent fuel and transmute them into short-lived or stable radionuclides in currently operating light-water reactors (LWR), thus reducing the radiological toxicity of the nuclear waste stream. One of the challenges is to demonstrate that the burnup-dependent characteristic differences between Reactor-Grade Mixed Oxide (RG-MOX) fuel and RG-MOX fuel with MA Np-237 and Am 241 are minimal, particularly, the inert gas generation rate, such that the commercial MOX fuel experience base is applicable. Under the Advanced Fuel Cycle Initiative (AFCI), developmental fuel specimens in experimental assembly LWR-2 are being tested in the northwest (NW) I-24 irradiation position of the Advanced Test Reactor (ATR). The experiment uses MOX fuel test hardware, and contains capsules with MOX fuel consisting of mixed oxide manufactured fuel using reactor grade plutonium (RG-Pu) and mixed oxide manufactured fuel using RG-Pu with added Np/Am. This study will compare the fuel neutronics depletion characteristics of Case-1 RG-MOX and Case-2 RG-MOX with Np/Am.

  17. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    SciTech Connect (OSTI)

    Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy; Scaglione, John M.

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

  18. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    SciTech Connect (OSTI)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next Generation Safeguards Initiative (NGSI).

  19. Hot startup experience with electrometallurgical treatment of spent nuclear fuel

    SciTech Connect (OSTI)

    Benedict, R.W.; Lineberry, M.J.; McFarlane, H.F.; Rigg, R.H.

    1997-10-01

    The treatment of spent metal fuel from the EBR-II fast reactor commenced in June of 1996 at the Fuel Conditioning Facility on the Argonne-West site in Idaho, USA. During the first year of hot operations, 20 fuel assemblies entered processing and 6 low enrichment uranium product ingots were produced. Results are presented for the various process steps with decontamination factors achieved and equipment operational history reported.

  20. Recycling of nuclear spent fuel with AIROX processing

    SciTech Connect (OSTI)

    Majumdar, D.; Jahshan, S.N.; Allison, C.M.; Kuan, P.; Thomas, T.R.

    1992-12-01

    This report examines the concept of recycling light water reactor (LWR) fuel through use of a dry-processing technique known as the AIROX (Atomics International Reduction Oxidation) process. In this concept, the volatiles and the cladding from spent LWR fuel are separated from the fuel by the AIROX process. The fuel is then reenriched and made into new fuel pins with new cladding. The feasibility of the concept is studied from a technical and high level waste minimization perspective.

  1. Industrial co-generation through use of a medium BTU gas from biomass produced in a high throughput reactor

    SciTech Connect (OSTI)

    Feldmann, H.F.; Ball, D.A.; Paisley, M.A.

    1983-01-01

    A high-throughput gasification system has been developed for the steam gasification of woody biomass to produce a fuel gas with a heating value of 475 to 500 Btu/SCF without using oxygen. Recent developments have focused on the use of bark and sawdust as feedstocks in addition to wood chips and the testing of a new reactor concept, the so-called controlled turbulent zone (CTZ) reactor to increase gas production per unit of wood fed. Operating data from the original gasification system and the CTZ system are used to examine the preliminary economics of biomass gasification/gas turbine cogeneration systems. In addition, a ''generic'' pressurized oxygen-blown gasification system is evaluated. The economics of these gasification systems are compared with a conventional wood boiler/steam turbine cogeneration system.

  2. Thermoacoustic Thermometry for Nuclear Reactor Monitoring

    SciTech Connect (OSTI)

    James A. Smith; Dale K. Kotter; Steven L. Garrett; Randall A. Ali

    2013-06-01

    On Friday, March 11, 2011, at 2:46pm (Japan Standard Trme), the Tohoku region on the east coast of northern Japan experi­enced what would become known as the largest earthquake in the country's history at magnitude 9.0 on the Richter scale. The Fukushima Daiichi nuclear power plant suffered exten­sive and irreversible damage. Six operating units were at the site, each with a boiling water reactor. When the earthquake struck, three of the six reactors were operating and the others were in a periodic inspection outage phase. In one reactor, all of the fuel had been relocated to a spent fuel pool in the reactor building. The seismic acceleration caused by the earthquake brought the three operating units to an automatic shutdown. Since there was damage to the power transmission lines, the emergency diesel generators (EDG) were automat­ically started to ensure continued cooling of the reactors and spent fuel pools. The situation was under control until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 meters, which was three times taller than the sea wall of 5m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to five of the six reactors. The flooding also resulted in the loss of instrumentation that would have other­ wise been used to monitor and control the emergency. The ugly aftermath included high radiation exposure to operators at the nuclear power plants and early contamina­tion of food supplies and water within several restricted areas in Japan, where high radiation levels have rendered them un­safe for human habitation. While the rest of the story will remain a tragic history, it is this part of the series of unfortunate events that has inspired our research. It has indubitably highlighted the need for a novel sensor and instrumentation system that can withstand similar or worse conditions to avoid future catastrophe and assume damage prevention as quickly as possible. This is the question which we are attempting to answer: Is it possible to implement a self-powered sensor that could transmit data independently of electronic networks while taking advantage of the harsh operating environment of the nuclear reactor?

  3. Comparison of actinide production in traveling wave and pressurized water reactors

    SciTech Connect (OSTI)

    Osborne, A.G.; Smith, T.A.; Deinert, M.R.

    2013-07-01

    The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

  4. A Thermo-Mechanical Analysis for a Nozzle Header of a Once-Through Steam Generator Designed for an Integral Reactor

    SciTech Connect (OSTI)

    Kim, YongWan; Kim, Dong Ok; Lee, Jae Seon; Kim, Jong In; Zee, Sung Quun [Korea Atomic Energy Research Institute, PO Box 105, YuSong, Taejon, 305-600 (Korea, Republic of)

    2004-07-01

    Thermo-mechanical behavior of the nozzle header of a steam generator developed for an integral reactor was investigated using experimental and finite element methods. The nozzle feedwater header suffers from severe thermal transient loadings during the operation of the nuclear reactor. The nozzle header is exposed to the low temperature inlet feedwater and the high temperature outlet superheated steam. The other side of the nozzle header is in contacts with the high temperature primary coolant. This temperature gradient results in high thermal stresses in the nozzle header structure. The thermal transient loading has been simulated in the test loop. The thermo-hydraulic parameters of the primary and the secondary system were controlled according to the operation mode programmed in the computer. Strain gauges and thermocouples attached at the highly stressed region monitored the thermo-mechanical behavior of the nozzle header. In parallel to the experimental study, the transient behavior of the nozzle header was simulated utilizing a commercial finite element code. The fluid temperature and the pressure obtained from the test loop were used for the input of the finite element analysis. As a result of this investigation, the thermo-mechanical load carrying capacity of the developed steam generator nozzle header was proved numerically and experimentally. (authors)

  5. Spent-fuel dry-storage testing at E-MAD (March 1978-March 1982)

    SciTech Connect (OSTI)

    Unterzuber, R.; Milnes, R.D.; Marinkovich, B.A.; Kubancsek, G.M.

    1982-09-01

    From March 1978 through March 1982, spent fuel dry storage tests were conducted at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site to confirm that commercial reactor spent fuel could be encapsulated and passively stored in one or more interim dry storage cell concepts. These tests were: electrically heated drywell, isolated and adjacent drywell, concrete silo, fuel assembly internal temperature measurement, and air-cooled vault. This document presents the test data and results as well as results from supporting test operations (spent fuel calorimetry and canister gas sampling).

  6. Spent Fuel Test-Climax: technical measurements data management system description and data presentation

    SciTech Connect (OSTI)

    Carlson, R.C.

    1985-08-01

    The Spent Fuel Test-Climax (SFT-C) was located 420 m below surface in the Climax Stock granite on the Nevada Test Site. The test was conducted under the technical direction of the Lawrence Livermore National Laboratory (LLNL) as part of the Nevada Nuclear Waste Storage Investigations (NNWSI) for the US Department of Energy. Eleven canisters of spent nuclear reactor fuel were emplaced, along with six electrical simulators, in April-May 1980. The spent fuel canisters were retrieved and the electrical simulators de-energized in March-April 1983. During the test, just over 1000 MW-hr of thermal energy was deposited in the site, causing temperature changes 100{sup 0}C near the canisters, and about 5{sup 0} in the tunnels. More than 900 channels of geotechnical, seismological, and test status data were recorded on nearly continuous basis for about 3-1/2 years, ending in September 1983. Most geotechnical instrumentation was known to be temperature sensitive, and thus would require temperature compensation before interpretation. Accordingly, a 10-in. reel of digital tape was off-loaded and shipped to Livermore every 4 to 8 weeks, where the data were verified, organized into 45 one-million-word files, and temperature corrected. The purpose of this report is to document the receipt and processing of the data by LLNL Livermore personnel, present facts about the history of the instruments which may be important to the interpretation of the data, present the data themselves in graphical form for each instrument over its operating lifetime, document the forms and locations in which the data will be archived, and offer the data to the geotechnical community for future use in understanding and predicting the effects of the storage of heat-generating waste in hard rocks such as granite.

  7. Neutronic Assessment of Transmutation Target Compositions in Heterogeneous Sodium Fast Reactor Geometries

    SciTech Connect (OSTI)

    Samuel E. Bays; Rodolfo M. Ferrer; Michael A. Pope; Benoit Forget; Mehdi Asgari

    2008-02-01

    The sodium fast reactor is under consideration for consuming the transuranic waste in the spent nuclear fuel generated by light water reactors. This work is concerned with specialized target assemblies for an oxide-fueled sodium fast reactor that are designed exclusively for burning the americium and higher mass actinide component of light water reactor spent nuclear fuel (SNF). The associated gamma and neutron radioactivity, as well as thermal heat, associated with decay of these actinides may significantly complicate fuel handling and fabrication of recycled fast reactor fuel. The objective of using targets is to isolate in a smaller number of assemblies these concentrations of higher actinides, thus reducing the volume of fuel having more rigorous handling requirements or a more complicated fabrication process. This is in contrast to homogeneous recycle where all recycled actinides are distributed among all fuel assemblies. Several heterogeneous core geometries were evaluated to determine the fewest target assemblies required to burn these actinides without violating a set of established fuel performance criteria. The DIF3D/REBUS code from Argonne National Laboratory was used to perform the core physics and accompanying fuel cycle calculations in support of this work. Using the REBUS code, each core design was evaluated at the equilibrium cycle condition.

  8. Systems for the Intermodal Routing of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Peterson, Steven K; Liu, Cheng

    2015-01-01

    The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable system for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of prioritization constraints and modifiers to determine route selection. The limitations of the current model and future directions for development are discussed, including the current state of information on possible intermodal transfer locations for spent fuel.

  9. Assessment of the safety of spent fuel transportation in urban environs

    SciTech Connect (OSTI)

    Sandoval, R.P.; Weber, J.P.; Levine, H.S.; Romig, A.D.; Johnson, J.D.; Luna, R.E.; Newton, G.J.; Wong, B.A.; Marshall, R.W. Jr.; Alvarez, J.L.

    1983-06-01

    The results of a program to provide an experimental data base for estimating the radiological consequences from a hypothetical sabotage attack on a light-water-reactor spent fuel shipping cask in a densely populated area are presented. The results of subscale and full-scale experiments in conjunction with an analytical modeling study are described. The experimental data were used as input to a reactor-safety consequence model to predict radiological health consequences resulting from a hypothetical sabotage attack on a spent-fuel shipping cask in the Manhattan borough of New York City. The results of these calculations are presented.

  10. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  11. Spent-fuel-storage alternatives

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  12. Shipper/receiver difference verification of spent fuel by use of PDET

    SciTech Connect (OSTI)

    Ham, Y. S.; Sitaraman, S.

    2011-07-01

    Spent fuel storage pools in most countries are rapidly approaching their design limits with the discharge of over 10,000 metric tons of heavy metal from global reactors. Countries like UK, France or Japan have adopted a closed fuel cycle by reprocessing spent fuel and recycling MOX fuel while many other countries opted for above ground interim dry storage for their spent fuel management strategy. Some countries like Finland and Sweden are already well on the way to setting up a conditioning plant and a deep geological repository for spent fuel. For all these situations, shipments of spent fuel are needed and the number of these shipments is expected to increase significantly. Although shipper/receiver difference (SRD) verification measurements are needed by IAEA when the recipient facility receives spent fuel, these are not being practiced to the level that IAEA has desired due to lack of a credible measurement methodology and instrument that can reliably perform these measurements to verify non-diversion of spent fuel during shipment and confirm facility operator declarations on the spent fuel. In this paper, we describe a new safeguards method and an associated instrument, Partial Defect Tester (PDET), which can detect pin diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies in an in-situ condition. The PDET uses multiple tiny neutron and gamma detectors in the form of a cluster and a simple, yet highly precise, gravity-driven system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly. The method takes advantage of the PWR fuel design which contains multiple guide tubes which can be accessed from the top. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. Our simulation study as well as validation measurements indicated that the ratio of the gamma signal to the thermal neutron signal at each detector location normalized to the peak ratio of all the detector locations gives a unique signature that is sensitive to missing pins. The signature is principally dependent on the geometry of the detector locations, and little sensitive to enrichment or burn-up variations. A small variation in the fuel bundle, such as a few missing pins, changes the shape of the signature to enable detection. After verification of the non-diversion of spent fuel pins, the neutron signal and gamma signal are subsequently used to verify the consistency of the operator declaration on the fuel burn-up and cooling time. (authors)

  13. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    SciTech Connect (OSTI)

    James A. Smith; Dale K. Kotter; Randall A. Ali; Steven L . Garrett

    2013-10-01

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  14. Metallic Fast Reactor Fuel Fabrication for Global Nuclear Energy Partnership

    SciTech Connect (OSTI)

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter

    2009-07-01

    Fast reactors are once again being considered for nuclear power generation, in addition to transmutation of long-lived fission products resident in spent nuclear fuels. This re-consideration follows with intense developmental programs for both fuel and reactor design. One of the two leading candidates for next generation fast reactor fuel is metal alloys, resulting primarily from the successes achieved in the 1960s to early 1990s with both the experimental breeding reactor-II and the fast flux test facility. The goal of the current program is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional, fast-spectrum nuclear fuel while destroying recycled actinides, thereby closing the nuclear fuel cycle. In order to meet this goal, the program must develop efficient and safe fuel fabrication processes designed for remote operation. This paper provides an overview of advanced casting processes investigated in the past, and the development of a gaseous diffusion calculation that demonstrates how straightforward process parameter modification can mitigate the loss of volatile minor actinides in the metal alloy melt.

  15. Annual report, FY 1979 Spent fuel and fuel pool component integrity.

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

    1980-05-01

    International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-..mu..m) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report.

  16. Development and Validation of Temperature Dependent Thermal Neutron Scattering Laws for Applications and Safety Implications in Generation IV Reactor Designs

    SciTech Connect (OSTI)

    Ayman Hawari

    2008-06-20

    The overall obljectives of this project are to critically review the currently used thermal neutron scattering laws for various moderators as a function of temperature, select as well documented and representative set of experimental data sensitive to the neutron spectra to generate a data base of benchmarks, update models and models parameters by introducing new developments in thermalization theory and condensed matter physics into various computational approaches in establishing the scattering laws, benchmark the results against the experimentatl set. In the case of graphite, a validation experiment is performed by observing nutron slowing down as a function of temperatures equal to or greater than room temperature.

  17. Design criteria for an independent spent fuel storage installation (water pool type)

    SciTech Connect (OSTI)

    Not Available

    1981-01-01

    This standard is intended to be used by those involved in the ownership and operation of an Independent Spent Fuel Storage Installation (ISFSI) in specifying the design requirements and by the designer in meeting the minimum design requirements of such installations. This standard continues the set of American National Standards on spent fuel storage design. Similar standards are: Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations, N210-1976 (ANS-57.2); Design Objectives for Highly Radioactive Solid Material Handling and Storage Facilities in a Reprocessing Plant, ANSI N305-1975; and Guidelines for Evaluating Site-Related Parameters for an Independent Spent Fuel Storage Installation, ANSI/ANS-2.19-1981.

  18. Advanced Nuclear Reactors | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key

  19. An economic analysis of a light and heavy water moderated reactor synergy: burning americium using recycled uranium

    SciTech Connect (OSTI)

    Wojtaszek, D.; Edwards, G.

    2013-07-01

    An economic analysis is presented for a proposed synergistic system between 2 nuclear utilities, one operating light water reactors (LWR) and another running a fleet of heavy water moderated reactors (HWR). Americium is partitioned from LWR spent nuclear fuel (SNF) to be transmuted in HWRs, with a consequent averted disposal cost to the LWR operator. In return, reprocessed uranium (RU) is supplied to the HWRs in sufficient quantities to support their operation both as power generators and americium burners. Two simplifying assumptions have been made. First, the economic value of RU is a linear function of the cost of fresh natural uranium (NU), and secondly, plutonium recycling for a third utility running a mixed oxide (MOX) fuelled reactor fleet has been already taking place, so that the extra cost of americium recycling is manageable. We conclude that, in order for this scenario to be economically attractive to the LWR operator, the averted disposal cost due to partitioning americium from LWR spent fuel must exceed 214 dollars per kg, comparable to estimates of the permanent disposal cost of the high level waste (HLW) from reprocessing spent LWR fuel. (authors)

  20. AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

    SciTech Connect (OSTI)

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

    In June 2009 Romania successfully completed the world’s first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments were transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.

  1. Evaluation of Radiation Impacts of Spent Nuclear Fuel Storage (SNFS-2) of Chernobyl NPP - 13495

    SciTech Connect (OSTI)

    Paskevych, Sergiy; Batiy, Valiriy; Sizov, Andriy; Schmieman, Eric

    2013-07-01

    Radiation effects are estimated for the operation of a new dry storage facility for spent nuclear fuel (SNFS-2) of Chernobyl NPP RBMK reactors. It is shown that radiation exposure during normal operation, design and beyond design basis accidents are minor and meet the criteria for safe use of radiation and nuclear facilities in Ukraine. (authors)

  2. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures.

  3. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented.

  4. Thermoelectric powered wireless sensors for spent fuel monitoring

    SciTech Connect (OSTI)

    Carstens, T.; Corradini, M.; Blanchard, J.; Ma, Z.

    2011-07-01

    This paper describes using thermoelectric generators to power wireless sensors to monitor spent nuclear fuel during dry-cask storage. OrigenArp was used to determine the decay heat of the spent fuel at different times during the service life of the dry-cask. The Engineering Equation Solver computer program modeled the temperatures inside the spent fuel storage facility during its service life. The temperature distribution in a thermoelectric generator and heat sink was calculated using the computer program Finite Element Heat Transfer. From these temperature distributions the power produced by the thermoelectric generator was determined as a function of the service life of the dry-cask. In addition, an estimation of the path loss experienced by the wireless signal can be made based on materials and thickness of the structure. Once the path loss is known, the transmission power and thermoelectric generator power requirements can be determined. This analysis estimates that a thermoelectric generator can produce enough power for a sensor to function and transmit data from inside the dry-cask throughout its service life. (authors)

  5. Fast reactors and nuclear nonproliferation

    SciTech Connect (OSTI)

    Avrorin, E.N.; Rachkov, V.I.; Chebeskov, A.N.

    2013-07-01

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  6. Safeguards-by-Design: Guidance for Independent Spent Fuel Dry Storage Installations (ISFSI)

    SciTech Connect (OSTI)

    Trond Bjornard; Philip C. Durst

    2012-05-01

    This document summarizes the requirements and best practices for implementing international nuclear safeguards at independent spent fuel storage installations (ISFSIs), also known as Away-from- Reactor (AFR) storage facilities. These installations may provide wet or dry storage of spent fuel, although the safeguards guidance herein focuses on dry storage facilities. In principle, the safeguards guidance applies to both wet and dry storage. The reason for focusing on dry independent spent fuel storage installations is that this is one of the fastest growing nuclear installations worldwide. Independent spent fuel storage installations are typically outside of the safeguards nuclear material balance area (MBA) of the reactor. They may be located on the reactor site, but are generally considered by the International Atomic Energy Agency (IAEA) and the State Regulator/SSAC to be a separate facility. The need for this guidance is becoming increasingly urgent as more and more nuclear power plants move their spent fuel from resident spent fuel ponds to independent spent fuel storage installations. The safeguards requirements and best practices described herein are also relevant to the design and construction of regional independent spent fuel storage installations that nuclear power plant operators are starting to consider in the absence of a national long-term geological spent fuel repository. The following document has been prepared in support of two of the three foundational pillars for implementing Safeguards-by-Design (SBD). These are: i) defining the relevant safeguards requirements, and ii) defining the best practices for meeting the requirements. This document was prepared with the design of the latest independent dry spent fuel storage installations in mind and was prepared specifically as an aid for designers of commercial nuclear facilities to help them understand the relevant international requirements that follow from a country’s safeguards agreement with the IAEA. If these requirements are understood at the earliest stages of facility design, it will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards, and will help the IAEA implement nuclear safeguards worldwide, especially in countries building their first nuclear facilities. It is also hoped that this guidance document will promote discussion between the IAEA, State Regulator/SSAC, Project Design Team, and Facility Owner/Operator at an early stage to ensure that new ISFSIs will be effectively and efficiently safeguarded. This is intended to be a living document, since the international nuclear safeguards requirements may be subject to revision over time. More importantly, the practices by which the requirements are met are continuously modernized by the IAEA and facility operators for greater efficiency and cost effectiveness. As these improvements are made, it is recommended that the subject guidance document be updated and revised accordingly.

  7. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, John P.

    1992-01-01

    A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  8. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, J.P.

    1992-03-17

    A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  9. Training implementation matrix, Spent Nuclear Fuel Project (SNFP)

    SciTech Connect (OSTI)

    EATON, G.L.

    2000-06-08

    This Training Implementation Matrix (TIM) describes how the Spent Nuclear Fuel Project (SNFP) implements the requirements of DOE Order 5480.20A, Personnel Selection, Qualification, and Training Requirements for Reactor and Non-Reactor Nuclear Facilities. The TIM defines the application of the selection, qualification, and training requirements in DOE Order 5480.20A at the SNFP. The TIM also describes the organization, planning, and administration of the SNFP training and qualification program(s) for which DOE Order 5480.20A applies. Also included is suitable justification for exceptions taken to any requirements contained in DOE Order 5480.20A. The goal of the SNFP training and qualification program is to ensure employees are capable of performing their jobs safely and efficiently.

  10. A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility

    SciTech Connect (OSTI)

    S. Khericha

    2010-12-01

    The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3 to 5 years to construct.

  11. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

  12. Reforming of fuel inside fuel cell generator

    DOE Patents [OSTI]

    Grimble, Ralph E. (Finleyville, PA)

    1988-01-01

    Disclosed is an improved method of reforming a gaseous reformable fuel within a solid oxide fuel cell generator, wherein the solid oxide fuel cell generator has a plurality of individual fuel cells in a refractory container, the fuel cells generating a partially spent fuel stream and a partially spent oxidant stream. The partially spent fuel stream is divided into two streams, spent fuel stream I and spent fuel stream II. Spent fuel stream I is burned with the partially spent oxidant stream inside the refractory container to produce an exhaust stream. The exhaust stream is divided into two streams, exhaust stream I and exhaust stream II, and exhaust stream I is vented. Exhaust stream II is mixed with spent fuel stream II to form a recycle stream. The recycle stream is mixed with the gaseous reformable fuel within the refractory container to form a fuel stream which is supplied to the fuel cells. Also disclosed is an improved apparatus which permits the reforming of a reformable gaseous fuel within such a solid oxide fuel cell generator. The apparatus comprises a mixing chamber within the refractory container, means for diverting a portion of the partially spent fuel stream to the mixing chamber, means for diverting a portion of exhaust gas to the mixing chamber where it is mixed with the portion of the partially spent fuel stream to form a recycle stream, means for injecting the reformable gaseous fuel into the recycle stream, and means for circulating the recycle stream back to the fuel cells.

  13. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect (OSTI)

    Douglas Morrell

    2011-03-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  14. Spent fuel performance data: An analysis of data relevant to the NNWSI Project

    SciTech Connect (OSTI)

    Oversby, V.M.; Shaw, H.F.

    1987-08-01

    This paper summarizes the physical and chemical properties of spent light water reactor fuel that might influence its performance as a waste form under geologic disposal conditions at Yucca Mountain, Nevada. Results obtained on the dissolution testing of spent fuel conducted by the NNWSI Project are presented and discussed. Work published by other programs, in particular those of Canada and Sweden, are reviewed and compared with the NNWSI testing results. An attempt is made to relate all of the results to a common basis of presentation and to rationalize apparent conflicts between sets of results obtained under different experimental conditions.

  15. Electrorefining Experience For Pyrochemical Reprocessing of Spent EBR-II Driver Fuel

    SciTech Connect (OSTI)

    S. X. Li; T. A. Johnson; B. R. Westphal; K. M. Goff; R. W. Benedict

    2005-10-01

    Pyrochemical processing has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor-II (EBR-II) at Idaho National Laboratory since 1996. This report summarizes technical advancements made in electrorefining of spent EBR-II driver fuel in the Mk-IV electrorefiner since the pyrochemical processing was integrated into the AFCI program in 2002. The significant advancements include improving uranium dissolution and noble metal retention from chopped fuel segments, increasing cathode current efficiency, and achieving co-collection of zirconium along with uranium from the cadmium pool.

  16. Actinide Burning in CANDU Reactors

    SciTech Connect (OSTI)

    Hyland, B.; Dyck, G.R.

    2007-07-01

    Actinide burning in CANDU reactors has been studied as a method of reducing the actinide content of spent nuclear fuel from light water reactors, and thereby decreasing the associated long term decay heat load. In this work simulations were performed of actinides mixed with natural uranium to form a mixed oxide (MOX) fuel, and also mixed with silicon carbide to form an inert matrix (IMF) fuel. Both of these fuels were taken to a higher burnup than has previously been studied. The total transuranic element destruction calculated was 40% for the MOX fuel and 71% for the IMF. (authors)

  17. On the RA research reactor fuel management problems

    SciTech Connect (OSTI)

    Matausek, M.V.; Marinkovic, N.

    1997-12-01

    After 25 yr of operation, the Soviet-origin 6.5-MW heavy water RA research reactor was shut down in 1984. Basic facts about RA reactor operation, aging, reconstruction, and spent-fuel disposal have been presented and discussed in earlier papers. The following paragraphs present recent activities and results related to important fuel management problems.

  18. Characteristics of fuel crud and its impact on storage, handling, and shipment of spent fuel. [Fuel crud

    SciTech Connect (OSTI)

    Hazelton, R.F.

    1987-09-01

    Corrosion products, called ''crud,'' form on out-of-reactor surfaces of nuclear reactor systems and are transported by reactor coolant to the core, where they deposit on external fuel-rod cladding surfaces and are activated by nuclear reactions. After discharge of spent fuel from a reactor, spallation of radioactive crud from the fuel rods could impact wet or dry storage operations, handling (including rod consolidation), and shipping. It is the purpose of this report to review earlier (1970s) and more recent (1980s) literature relating to crud, its characteristics, and any impact it has had on actual operations. Crud characteristics vary from reactor type to reactor type, reactor to reactor, fuel assembly to fuel assembly in a reactor, circumferentially and axially in an assembly, and from cycle to cycle for a specific facility. To characterize crud of pressurized-water (PWRs) and boiling-water reactors (BWRs), published information was reviewed on appearance, chemical composition, areal density and thickness, structure, adhesive strength, particle size, and radioactivity. Information was also collected on experience with crud during spent fuel wet storage, rod consolidation, transportation, and dry storage. From experience with wet storage, rod consolidation, transportation, and dry storage, it appears crud spallation can be managed effectively, posing no significant radiological problems. 44 refs., 11 figs.

  19. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect (OSTI)

    Shropshire, D.E.; Herring, J.S.

    2004-10-03

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim storage, packaging, transportation, waste forms, waste treatment, decontamination and decommissioning issues; and low-level waste (LLW) and high-level waste (HLW) disposal.

  20. Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences

    SciTech Connect (OSTI)

    Bailey, W.J.

    1990-02-01

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs.

  1. Spent Nuclear Fuel (SNF) Project Execution Plan

    SciTech Connect (OSTI)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  2. Automated Characterization of Spent Fuel through the Multi-Isotope Process (MIP) Monitor

    SciTech Connect (OSTI)

    Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M.

    2012-07-31

    This research developed an algorithm for characterizing spent nuclear fuel (SNF) samples based on simulated gamma spectra. The gamma spectra for a variety of light water reactor fuels typical of those found in the United States were simulated. Fuel nuclide concentrations were simulated in ORIGEN-ARP for 1296 fuel samples with a variety of reactor designs, initial enrichments, burn ups, and cooling times. The results of the ORIGEN-ARP simulation were then input to SYNTH to simulate the gamma spectrum for each sample. These spectra were evaluated with partial least squares (PLS)-based multivariate analysis methods to characterize the fuel according to reactor type (pressurized or boiling water reactor), enrichment, burn up, and cooling time. Characterizing some of the features in series by using previously estimated features in the prediction greatly improves the performance. By first classifying the spent fuel reactor type and then using type-specific models, the prediction error for enrichment, burn up, and cooling time improved by a factor of two to four. For some features, the prediction was further improved by including additional information, such as including the predicted burn up in the estimation of cooling time. The optimal prediction flow was determined based on the simulated data. A PLS discriminate analysis model was developed which perfectly classified SNF reactor type. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE.

  3. FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

    SciTech Connect (OSTI)

    Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.; Livingston, James V.

    2014-03-31

    This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset will be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.

  4. MELCOR Model of the Spent Fuel Pool of Fukushima Dai-ichi Unit 4

    SciTech Connect (OSTI)

    Carbajo, Juan J

    2012-01-01

    Unit 4 of the Fukushima Dai-ichi Nuclear Power Plant suffered a hydrogen explosion at 6:00 am on March 15, 2011, exactly 3.64 days after the earthquake hit the plant and the off-site power was lost. The earthquake occurred on March 11 at 2:47 pm. Since the reactor of this Unit 4 was defueled on November 29, 2010, and all its fuel was stored in the spent fuel pool (SFP4), it was first believed that the explosion was caused by hydrogen generated by the spent fuel, in particular, by the recently discharged core. The hypothetical scenario was: power was lost, cooling to the SFP4 water was lost, pool water heated/boiled, water level decreased, fuel was uncovered, hot Zircaloy reacted with steam, hydrogen was generated and accumulated above the pool, and the explosion occurred. Recent analyses of the radioisotopes present in the water of the SFP4 and underwater video indicated that this scenario did not occur - the fuel in this pool was not damaged and was never uncovered the hydrogen of the explosion was apparently generated in Unit 3 and transported through exhaust ducts that shared the same chimney with Unit 4. This paper will try to answer the following questions: Could that hypothetical scenario in the SFP4 had occurred? Could the spent fuel in the SPF4 generate enough hydrogen to produce the explosion that occurred 3.64 days after the earthquake? Given the magnitude of the explosion, it was estimated that at least 150 kg of hydrogen had to be generated. As part of the investigations of this accident, MELCOR models of the SFP4 were prepared and a series of calculations were completed. The latest version of MELCOR, version 2.1 (Ref. 1), was employed in these calculations. The spent fuel pool option for BWR fuel was selected in MELCOR. The MELCOR model of the SFP4 consists of a total of 1535 fuel assemblies out of which 548 assemblies are from the core defueled on Nov. 29, 2010, 783 assemblies are older assemblies, and 204 are new/fresh assemblies. The total decay heat of the fuel in the pool was, at the time of the accident, 2.284 MWt, of which 1.872 MWt were from the 548 assemblies of the last core discharged and 0.412 MWt were from the older 783 assemblies. These decay heat values were calculated at Oak Ridge National Laboratory using the ORIGEN2.2 code (Ref. 2) - they agree with values reported elsewhere (Ref. 3). The pool dimensions are 9.9 m x 12.2 m x 11.8 m (height), and with the water level at 11.5 m, the pool volume is 1389 m3, of which only 1240 m3 is water, as some volume is taken by the fuel and by the fuel racks. The initial water temperature of the SFP4 was assumed to be 301 K. The fuel racks are made of an aluminum alloy but are modeled in MELCOR with stainless steel and B4C. MELCOR calculations were completed for different initial water levels: 11.5 m (pool almost full, water is only 0.3 m below the top rim), 4.4577 m (top of the racks), 4.2 m, and 4.026 m (top of the active fuel). A calculation was also completed for a rapid loss of water due to a leak at the bottom of the pool, with the fuel rapidly uncovered and oxidized in air. Results of these calculations are shown in the enclosed Table I. The calculation with the initial water level at 11.5 m (full pool) takes 11 days for the water to boil down to the top of the fuel racks, 11.5 days for the fuel to be uncovered, 14.65 days to generate 150 kg of hydrogen and 19 days for the pool to be completely dry. The calculation with the initial water level at 4.4577 m, takes 1.1 days to uncover the fuel and 4.17 days to generate 150 kg of hydrogen. The calculation with the initial water level at 4.02 m takes 3.63 days to generate 150 kg of hydrogen this is exactly the time when the actual explosion occurred in Unit 4. Finally, fuel oxidation in air after the pool drained the water in 20 minutes, generates only 10 kg of hydrogen this is because very little steam is available and Zircaloy (Zr) oxidation with the oxygen of the air does not generate hydrogen. MELCOR calculated water levels and hydrogen generated in the SFP4 as a function of time for initial water levels of 4.457 m, 4.2 m and 4.02 m are shown in Figs. 1 and 2. Water levels increase at the beginning due to the expansion of the water during the heat-up from 301 K to 373 K. Boiling occurs after the water temperature reaches 373 K. The total amount of hydrogen generated is ~2000 kg, this amount includes hydrogen generated from Zr, which is the largest amount (~1580 kg), from stainless steel (~360 kg), and from B4C (~60 kg). In theory, it is possible to generate up to 3.4 kg of hydrogen per assembly (from oxidation of Zr in the fuel cladding and box), or a total of 4,525 kg from the hot 1331 assemblies stored in the SFP4. The hydrogen generated from oxidation of steel and B4C will be additional. So the answers to the questions are YES according to these MELCOR calculations, enough hydrogen (150 kg) could be generated in the SFP4 3.64 days after the earthquake to produce ...

  5. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2010-05-05

    The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

  6. PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

    SciTech Connect (OSTI)

    S. T. Khericha

    2007-04-01

    The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.

  7. Validation of SCALE (SAS2H) isotopic predictions for BWR spent fuel

    SciTech Connect (OSTI)

    Hermann, O.W.; DeHart, M.D.

    1998-09-01

    Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system. The SAS2H sequence uses transport methods combined with the depletion and decay capabilities of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup history. Results of these calculations are compared with chemical assay measurements of spent fuel inventories for each sample. Results show reasonable agreement between measured and predicted isotopic concentrations for important actinides; however, little data are available for most fission products considered to be important for spent fuel concerns (e.g., burnup credit, shielding, source-term calculations, etc.). This work is a follow-up to earlier works that studied the ability to predict spent fuel compositions in pressurized-water-reactor (PWR) fuel pins. Biases and uncertainties associated with BWR isotopic predictions are found to be larger than those of PWR calculations. Such behavior is expected, as the operation of a BWR is significantly more complex than that of a PWR plant, and in general the design of a BWR has a more heterogeneous configuration than that of a PWR. Nevertheless, this work shows that the simple models employed using SAS2H to represent such complexities result in agreement to within 5% (and often less than 1%) or less for most nuclides important for spent fuel applications. On the other hand, however, the set of fuel samples analyzed represent a small subset of the BWR fuel population, and results reported herein may not be representative of the full population of BWR spent fuel.

  8. Performance Comparison of Metallic, Actinide Burning Fuel in Lead-Bismuth and Sodium Cooled Fast Reactors

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-04-01

    Various methods have been proposed to “incinerate” or “transmutate” the current inventory of trans-uranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non-fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years.

  9. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Volume 1, Appendix B: Idaho National Engineering Laboratory Spent Nuclear Fuel Management Program

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    The US Department of Energy (DOE) has prepared this report to assist its management in making two decisions. The first decision, which is programmatic, is to determine the management program for DOE spent nuclear fuel. The second decision is on the future direction of environmental restoration, waste management, and spent nuclear fuel management activities at the Idaho National Engineering Laboratory. Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent nuclear fuel management on the quality of the human and natural environment for planning years 1995 through 2035. DOE has derived the information and analysis results in Volume 1 from several site-specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes environmental impacts for various environmental restoration, waste management, and spent nuclear fuel management alternatives for planning years 1995 through 2005. This Appendix B to Volume 1 considers the impacts on the INEL environment of the implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the management of DOE spent nuclear fuel including naval-type fuel.

  10. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs, Draft Environmental Impact Statement. Volume 1, Appendix D: Part A, Naval Spent Nuclear Fuel Management

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    Volume 1 to the Department of Energy`s Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Management Programs Environmental Impact Statement evaluates a range of alternatives for managing naval spent nuclear fuel expected to be removed from US Navy nuclear-powered vessels and prototype reactors through the year 2035. The Environmental Impact Statement (EIS) considers a range of alternatives for examining and storing naval spent nuclear fuel, including alternatives that terminate examination and involve storage close to the refueling or defueling site. The EIS covers the potential environmental impacts of each alternative, as well as cost impacts and impacts to the Naval Nuclear Propulsion Program mission. This Appendix covers aspects of the alternatives that involve managing naval spent nuclear fuel at four naval shipyards and the Naval Nuclear Propulsion Program Kesselring Site in West Milton, New York. This Appendix also covers the impacts of alternatives that involve examining naval spent nuclear fuel at the Expended Core Facility in Idaho and the potential impacts of constructing and operating an inspection facility at any of the Department of Energy (DOE) facilities considered in the EIS. This Appendix also considers the impacts of the alternative involving limited spent nuclear fuel examinations at Puget Sound Naval Shipyard. This Appendix does not address the impacts associated with storing naval spent nuclear fuel after it has been inspected and transferred to DOE facilities. These impacts are addressed in separate appendices for each DOE site.

  11. Deep Borehole Disposal of Spent Fuel. (Conference) | SciTech...

    Office of Scientific and Technical Information (OSTI)

    Deep Borehole Disposal of Spent Fuel. Citation Details In-Document Search Title: Deep Borehole Disposal of Spent Fuel. Abstract not provided. Authors: Brady, Patrick V. Publication...

  12. Activities Related to Storage of Spent Nuclear Fuel | Department...

    Office of Environmental Management (EM)

    Related to Storage of Spent Nuclear Fuel More Documents & Publications Nuclear Regulatory Commission Fifth National Report for the Joint Convention on the Safety of Spent...

  13. National Report Joint Convention on the Safety of Spent Fuel...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management National Report Joint Convention on the Safety of Spent Fuel Management...

  14. Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy Lessons Learned Report, NNSA, Feb 2010 Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition...

  15. Current Status of the Spent Nuclear Fuel Management Program in...

    Office of Scientific and Technical Information (OSTI)

    Current Status of the Spent Nuclear Fuel Management Program in the United States. Citation Details In-Document Search Title: Current Status of the Spent Nuclear Fuel Management...

  16. Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

    SciTech Connect (OSTI)

    N /A

    2000-04-14

    The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.

  17. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    SciTech Connect (OSTI)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  18. CORAL: a stepping stone for establishing the Indian fast reactor fuel reprocessing technology

    SciTech Connect (OSTI)

    Venkataraman, M.; Natarajan, R.; Raj, Baldev

    2007-07-01

    The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR) spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)

  19. Electrometallurgical treatment of oxide spent fuel - engineering-scale development.

    SciTech Connect (OSTI)

    Karell, E. J.

    1998-04-22

    Argonne National Laboratory (ANL) has developed the electrometallurgical treatment process for conditioning various Department of Energy (DOE) spent fuel types for long-term storage or disposal. This process uses electrorefining to separate the constituents of spent fuel into three product streams: metallic uranium, a metal waste form containing the cladding and noble metal fission products, and a ceramic waste form containing the transuranics, and rare earth, alkali, and alkaline earth fission products. While metallic fuels can be directly introduced into the electrorefiner, the actinide components of oxide fuels must first be reduced to the metallic form. The Chemical Technology Division of AFT has developed a process to reduce the actinide oxides that uses lithium at 650 C in the presence of molten LiCl, yielding the actinide metals and Li{sub 2}O. A significant amount of work has already been accomplished to investigate the basic chemistry of the lithium reduction process and to demonstrate its applicability to the treatment of light-water reactor- (LWR-) type spent fuel. The success of this work has led to conceptual plans to construct a pilot-scale oxide reduction facility at ANL's Idaho site. In support of the design effort, a series of laboratory- and engineering-scale experiments is being conducted using simulated fuel. These experiments have focused on the engineering issues associated with scaling-up the process and proving compatibility between the reduction and electrorefining steps. Specific areas of investigation included reduction reaction kinetics, evaluation of various fuel basket designs, and issues related to electrorefining the reduced product. This paper summarizes the results of these experiments and outlines plans for future work.

  20. An Innovative Three-Dimensional Heterogeneous Coarse-Mesh Transport Method for Advanced and Generation IV Reactor Core Analysis and Design

    SciTech Connect (OSTI)

    Farzad Rahnema

    2009-11-12

    This project has resulted in a highly efficient method that has been shown to provide accurate solutions to a variety of 2D and 3D reactor problems. The goal of this project was to develop (1) an accurate and efficient three-dimensional whole-core neutronics method with the following features: based sollely on transport theory, does not require the use of cross-section homogenization, contains a highly accurate and self-consistent global flux reconstruction procedure, and is applicable to large, heterogeneous reactor models, and to (2) create new numerical benchmark problems for code cross-comparison.

  1. Spent nuclear fuel recycling with plasma reduction and etching

    DOE Patents [OSTI]

    Kim, Yong Ho

    2012-06-05

    A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF4 or NF3 as the fluorine sources instead of F2 or HF.

  2. Nuclear Forensics Attributing the Source of Spent Fuel Used in an RDD Event

    SciTech Connect (OSTI)

    M.R. Scott

    2005-06-01

    An RDD attack against the U.S. is something America needs to prepare against. If such an event occurs the ability to quickly identify the source of the radiological material used in an RDD would aid investigators in identifying the perpetrators. Spent fuel is one of the most dangerous possible radiological sources for an RDD. In this work, a forensics methodology was developed and implemented to attribute spent fuel to a source reactor. The specific attributes determined are the spent fuel burnup, age from discharge, reactor type, and initial fuel enrichment. It is shown that by analyzing the post-event material, these attributes can be determined with enough accuracy to be useful for investigators. The burnup can be found within a 5% accuracy, enrichment with a 2% accuracy, and age with a 10% accuracy. Reactor type can be determined if specific nuclides are measured. The methodology developed was implemented into a code call NEMASYS. NEMASYS is easy to use and it takes a minimum amount of time to learn its basic functions. It will process data within a few minutes and provide detailed information about the results and conclusions.

  3. Composition of Insoluble Residues Generated During Spent Fuel...

    Office of Scientific and Technical Information (OSTI)

    OSTI Identifier: 828959 Resource Type: Conference Resource Relation: Conference: Waste Management 2002 Symposium, Tucson, AZ (US), 02242002--02282002; Other...

  4. Spent Nuclear Fuel Project Safety Management Plan

    SciTech Connect (OSTI)

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities.

  5. Nuclear reactor characteristics and operational history

    Gasoline and Diesel Fuel Update (EIA)

    Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: TBD See also: Table 1. Capacity and Generation, Table 2. Ownership Data Table 3. Nuclear Reactor Characteristics and Operational History PDF XLS Plant Name Generator ID Type Reactor Supplier and Model Construction Start Grid Connection Original Expiration Date License Renewal Application License Renewal Issued Extended Expiration Arkansas Nuclear One 1

  6. Spent Fuel Reprocessing: More Value for Money Spent in a Geological Repository?

    SciTech Connect (OSTI)

    Kaplan, P.; Vinoche, R.; Devezeaux, J-G.; Bailly, F.

    2003-02-25

    Today, each utility or country operating nuclear power plants can select between two long-term spent fuel management policies: either, spent fuel is considered as waste to dispose of through direct disposal or, spent fuel is considered a resource of valuable material through reprocessing-recycling. Reading and listening to what is said in the nuclear community, we understand that most people consider that the choice of policy is, actually, a choice among two technical paths to handle spent fuel: direct disposal versus reprocessing. This very simple situation has been recently challenged by analysis coming from countries where both policies are on survey. For example, ONDRAF of Belgium published an interesting study showing that, economically speaking for final disposal, it is worth treating spent fuel rather than dispose of it as a whole, even if there is no possibility to recycle the valuable part of it. So, the question is raised: is there such a one-to-one link between long term spent fuel management political option and industrial option? The purpose of the presentation is to discuss the potential advantages and drawbacks of spent fuel treatment as an implementation of the policy that considers spent fuel as waste to dispose of. Based on technical considerations and industrial experience, we will study qualitatively, and quantitatively when possible, the different answers proposed by treatment to the main concerns of spent-fuel-as-a-whole geological disposal.

  7. Pyrochemical processing of DOE spent nuclear fuel

    SciTech Connect (OSTI)

    Laidler, J.J.

    1995-02-01

    A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or {open_quotes}pyroprocessing,{close_quotes} provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (>99.9%) separation of transuranics. The resultant waste forms from the pyroprocess, are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory.

  8. Economic Viability of Brewery Spent Grain as a Biofuel

    SciTech Connect (OSTI)

    Morrow, Charles

    2016-01-01

    This report summarizes an investigation into the technical feasibility and economic viability of use grain wastes from the beer brewing process as fuel to generate the heat needed in subsequent brewing process. The study finds that while use of spent grain as a biofuel is technically feasible, the economics are not attractive. Economic viability is limited by the underuse of capital equipment. The investment in heating equipment requires a higher utilization that the client brewer currently anticipates. It may be possible in the future that changing factors may swing the decision to a more positive one.

  9. Small Modular Reactors (SMRs) | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Reactor Technologies » Small Modular Reactors (SMRs) Small Modular Reactors (SMRs) NuScale Power Reactors. ©NuScale Power, LLC, All Rights Reserved NuScale Power Reactors. ©NuScale Power, LLC, All Rights Reserved Small Modular Reactors (SMRs) are nuclear power plants that are smaller in size (300 MWe or less) than current generation base load plants (1,000 MWe or higher). These smaller, compact designs are factory-fabricated reactors that can be transported by truck or rail to a nuclear

  10. Impacts of a high-burnup spent fuel on a geological disposal system design

    SciTech Connect (OSTI)

    Cho, D.K.; Lee, Y.; Lee, J.Y.; Choi, H.J.; Choi, J.W. [Korea Atomic Energy Research Institute, Yuseong-gu, Daejeon-city (Korea, Republic of)

    2007-07-01

    The influence of a burnup increase of a spent nuclear fuel on a deep geological disposal system was evaluated in this study. First, the impact of a burnup increase on each aspect related to thermal and nuclear safety concerns was quantified. And then, the tunnel length, excavation volume, and the raw materials for a cast insert, copper, bentonite, and backfill needed to constitute a disposal system were comprehensively analyzed based on the spent fuel inventory to generate 1 Terawatt-year (TWa), to establish the overall effects and consequences on a geological disposal. As a result, impact of a burnup increase on the criticality safety and radiation shielding was shown to be negligible. The disposal area, however, is considerably affected because of a higher thermal load. And, it is reasonable to use a canister such as the Korean Reference Disposal Canister (KDC-1) containing 4 spent fuels up to 50 GWD/MtU, and to use a canister containing 3 spent fuels beyond 50 GWD/MtU. Although a considerable increased, 33 % in the tunnel length and 30 % in the excavation volume, was observed as the burnup increases from 50 to 60 GWD/MtU, because a decrease in the canister needs can offset an increase in the excavation volume, it can be concluded that a burnup increase of a spent fuel is not a critical concern for a geological disposal of a spent fuel. (authors)

  11. Advanced Safeguards Approaches for New Fast Reactors

    SciTech Connect (OSTI)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  12. CONVECTION REACTOR

    DOE Patents [OSTI]

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  13. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  14. Challenges in the Development of Advanced Reactors

    SciTech Connect (OSTI)

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  15. Neutron Resonance Transmission Analysis (NRTA): A Nondestructive Assay Technique for the Next Generation Safeguards Initiative’s Plutonium Assay Challenge

    SciTech Connect (OSTI)

    J. W. Sterbentz; D. L. Chichester

    2010-12-01

    This is an end-of-year report for a project funded by the National Nuclear Security Administration's Office of Nuclear Safeguards (NA-241). The goal of this project is to investigate the feasibility of using Neutron Resonance Transmission Analysis (NRTA) to assay plutonium in commercial light-water-reactor spent fuel. This project is part of a larger research effort within the Next-Generation Safeguards Initiative (NGSI) to evaluate methods for assaying plutonium in spent fuel, the Plutonium Assay Challenge. The first-year goals for this project were modest and included: 1) developing a zero-order MCNP model for the NRTA technique, simulating data results presented in the literature, 2) completing a preliminary set of studies investigating important design and performance characteristics for the NRTA measurement technique, and 3) documentation of this work in an end of the year report (this report). Research teams at Los Alamos National Laboratory (LANL), Lawrence Berkeley National Laboratory (LBNL), Pacific Northwest National Laboratory (PNNL), and at several universities are also working to investigate plutonium assay methods for spent-fuel safeguards. While the NRTA technique is well proven in the scientific literature for assaying individual spent fuel pins, it is a newcomer to the current NGSI efforts studying Pu assay method techniques having just started in March 2010; several analytical techniques have been under investigation within this program for two to three years or more. This report summarizes a nine month period of work.

  16. The CASTOR-V/21 PWR spent-fuel storage cask: Testing and analyses: Interim report

    SciTech Connect (OSTI)

    Dziadosz, D.; Moore, E.V.; Creer, J.M.; McCann, R.A.; McKinnon, M.A.; Tanner, J.E.; Gilbert, E.R.; Goodman, R.L.; Schoonen, D.H.; Jensen, M.

    1986-11-01

    A performance test of a Gesellschaft fuer Nuklear Service CASTOR-V/21 pressurized water reactor (PWR) spent fuel storage cask was performed. The test was the first of a series of cask performance tests planned under a cooperative agreement between Virginia Power and the US Department of Energy. The performance test consisted of loading the CASTOR-V/21 cask with 21 PWR spent fuel assemblies from Virginia Power's Surry reactor. Cask surface and fuel assembly guide tube temperatures, and cask surface gamma and neutron dose rates were measured. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Limited spent fuel integrity data were also obtained. Results of the performance test indicate the CASTOR-V/21 cask exhibited exceptionally good heat transfer performance which exceeded design expectations. Peak cladding temperatures with helium and nitrogen backfills in a vertical cast orientation and with helium in a horizontal orientation were less than the allowable of 380/sup 0/C with a total cask heat load of 28 kW. Significant convection heat transfer was present in vertical nitrogen and helium test runs as indicated by peak temperatures occurring in the upper regions of the fuel assemblies. Pretest temperature predictions of the HYDRA heat transfer computer program were in good agreement with test data, and post-test predictions agreed exceptionally well (25/sup 0/C) with data. Cask surface gamma and neutron dose rates were measured to be less than the design goal of 200 mrem/h. Localized peaks as high as 163 mrem/h were measured on the side of the cask, but peak dose rates of <75 mrem/h can easily be achieved with minor refinements to the gamma shielding design. From both heat transfer and shielding perspectives, the CASTOR-V/21 cask can, with minor refinements, be effectively implemented at reactor sites and central storage facilities for safe storage of spent fuel.

  17. Electrochemical cell apparatus having axially distributed entry of a fuel-spent fuel mixture transverse to the cell lengths

    DOE Patents [OSTI]

    Reichner, P.; Dollard, W.J.

    1991-01-08

    An electrochemical apparatus is made having a generator section containing axially elongated electrochemical cells, a fresh gaseous feed fuel inlet, a gaseous feed oxidant inlet, and at least one gaseous spent fuel exit channel, where the spent fuel exit channel passes from the generator chamber to combine with the fresh feed fuel inlet at a mixing apparatus, reformable fuel mixture channel passes through the length of the generator chamber and connects with the mixing apparatus, that channel containing entry ports within the generator chamber, where the axis of the ports is transverse to the fuel electrode surfaces, where a catalytic reforming material is distributed near the reformable fuel mixture entry ports. 2 figures.

  18. An approach to determine a defensible spent fuel ratio.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Lindgren, Eric Richard

    2014-03-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO2), have been conducted in the interim to more definitively determine the source term from these postulated events. In all the previous studies, the postulated attack of greatest interest was by a conical shape charge (CSC) that focuses the explosive energy much more efficiently than bulk explosives. However, the validity of these large-scale results remain in question due to the lack of a defensible Spent Fuel Ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical DUO2 surrogate. Previous attempts to define the SFR have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Different researchers have suggested using SFR values of 3 to 5.6. Sound technical arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and dry storage of spent nuclear fuel. Currently, Oak Ridge National Laboratory (ORNL) is in possession of several samples of spent nuclear fuel (SNF) that were used in the original SFR studies in the 1980's and were intended for use in a modern effort at Sandia National Laboratories (SNL) in the 2000's. A portion of these samples are being used for a variety of research efforts. However, the entirety of SNF samples at ORNL is scheduled for disposition at the Waste Isolation Pilot Plant (WIPP) by approximately the end of 2015. If a defensible SFR is to be determined for use in storage and transportation security analyses, the need to begin this effort is urgent in order to secure the only known available SNF samples with a clearly defined path to disposal.

  19. Safe Advantage on Dry Interim Spent Nuclear Fuel Storage

    SciTech Connect (OSTI)

    Romanato, L.S.

    2008-07-01

    This paper aims to present the advantages of dry cask storage in comparison with the wet storage (cooling water pools) for SNF. When the nuclear fuel is removed from the core reactor, it is moved to a storage unit and it wait for a final destination. Generally, the spent nuclear fuel (SNF) remains inside water pools within the reactors facility for the radioactive activity decay. After some period of time in pools, SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing facilities, or still, wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet facilities, depending on the method adopted by the nuclear power plant or other plans of the country. Interim storage, up to 20 years ago, was exclusively wet and if the nuclear facility had to be decommissioned another storage solution had to be found. At the present time, after a preliminary cooling of the SNF elements inside the water pool, the elements can be stored in dry facilities. This kind of storage does not need complex radiation monitoring and it is safer then wet one. Casks, either concrete or metallic, are safer, especially on occurrence of earthquakes, like that occurred at Kashiwazaki-Kariwa nuclear power plant, in Japan on July 16, 2007. (authors)

  20. Thermal reclamation of spent blasting abrasive

    SciTech Connect (OSTI)

    Bryan, B.G. ); Thomas, W.; Adema, C. )

    1990-01-01

    Abrasive blasting media is used to remove anticorrosive and antifoulant coatings from the hulls and tanks of US Navy ships. The total production of paint-contaminated spent abrasives from the eight US. Navy shipyards ranges from 75,000 to 100,000 tons per year. Most of this spent abrasive is disposed in landfills. Organic paint binders and heavy metals are present in the spent abrasives in concentrations sufficient to classify them as hazardous wastes in some states. In an effort to avoid the rising costs an long-term environmental liability associated with landfilling this waste, the US Navy has investigated various methods of reclaiming spent abrasives for reuse in hull- and tank-blasting operations. This paper discusses the results of a research and development project conducted under the Navy's Hazardous Waste Minimization Program to test a fluidized-bed sloped-grid (FBSG) reclaimer to determine if it could be used to recycle spent abrasive. Thirty tons of abrasive were processed and a product meeting military specifications for new abrasives was reclaimed. Blasting performance was also comparable to new abrasives. 3 refs., 1 fig., 2 tabs.

  1. Advanced Burner Reactor Preliminary NEPA Data Study.

    SciTech Connect (OSTI)

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly documents the extensive evaluation which was performed on the anticipated environmental impacts of that plant. This source can be referenced in the open literature and is publicly available. The CRBRP design was also of a commercial demonstration plant size - 975 MWth - which falls in the middle of the range of ABR plant sizes being considered (250 MWth to 2000 MWth). At the time the project was cancelled, the CRBRP had progressed to the point of having completed the licensing application to the Nuclear Regulatory Commission (NRC) and was in the process of receiving NRC approval. Therefore, it was felt that [CRBRP, 1977] provides some of the best available data and information as input to the GNEP PEIS work. CRBRP was not the source of all the information in this document. It is also expected that the CRBRP data will be bounding from the standpoint of commodity usage because fast reactor vendors will develop designs which will focus on commodity and footprint reduction to reduce the overall cost per kilowatt electric compared with the CRBR plant. Other sources used for this datacall information package are explained throughout this document and in Appendix A. In particular, see Table A.1 for a summary of the data sources used to generate the datacall information.

  2. Small modular reactors (SMRs) such...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Small modular reactors (SMRs) such as the one illustrated in Figure 1 are being considered by the commercial nuclear power industry as an option for more distributed generation and...

  3. Apparatus for shearing spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Weil, Bradley S.; Metz, III, Curtis F.

    1980-01-01

    A method and apparatus are described for shearing spent nuclear fuel assemblies of the type comprising an array of fuel pins disposed within an outer metal shell or shroud. A spent fuel assembly is first compacted in a known manner and then incrementally sheared using fixed and movable shear blades having matched laterally projecting teeth which slidably intermesh to provide the desired shearing action. Incremental advancement of the fuel assembly after each shear cycle is limited to a distance corresponding to the lateral projection of the teeth to ensure fuel assembly breakup into small uniform segments which are amenable to remote chemical processing.

  4. Method for shearing spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Weil, Bradley S.; Watson, Clyde D.

    1977-01-01

    A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

  5. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    SciTech Connect (OSTI)

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

  6. Spent fuel accumulations and arisings in 'fuel user' states: implications for a world nuclear partnership

    SciTech Connect (OSTI)

    Lineberry, M.J.

    2007-07-01

    Consider the question: In a GNEP world, what are the implications of current spent fuel inventories and possible future accumulations in 'fuel user' states? The inventory today ({approx}95,000 MTHM) and the growth rate ({approx}4,000 MTHM/yr) in likely fuel user states is roughly twice that in the U.S., and thus the problem is substantial. The magnitude of the issue increases if, as is expected, fuel user states grow their nuclear capacity. For purposes of illustration, 2% annual growth assumed in this study. There is an implied imperative to remove spent fuel inventories from fuel user states in some reasonable time (if the spent fuel backlog cannot be cleared by 2100, what motivation is there for GNEP today?). Clearing the backlog requires massive reprocessing capacity, but that makes no economic or strategic sense until the advanced burner reactors are developed and deployed. Thus, burner reactor development is the pacing item with any practical GNEP schedule. With regard to economics, cost estimates for reprocessing (which drive key GNEP costs) are much too disparate today. The disparities must be lessened in order to permit rational decision-making. (author)

  7. Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

    SciTech Connect (OSTI)

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin

    2010-07-01

    Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the world’s first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment.

  8. FIELD-DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE

    SciTech Connect (OSTI)

    Berry, T.; Milliken, C.; Martinez-Rodriguez, M.; Hathcock, D.; Heitkamp, M.

    2012-09-12

    Methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of aqueous spent fuel storage basins and determine the oxide thickness on the spent fuel basin materials were developed to assess the corrosion potential of a basin. this assessment can then be used to determine the amount of time fuel has spent in a storage basin to ascertain if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations and assist in evaluating general storage basin operations. The test kit was developed based on the identification of key physical, chemical and microbiological parameters identified using a review of the scientific and basin operations literature. The parameters were used to design bench scale test cells for additional corrosion analyses, and then tools were purchased to analyze the key parameters. The tools were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The sampling kit consisted of a total organic carbon analyzer, an YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization confirmed that the L Area basin is a well operated facility with low corrosion potential.

  9. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  10. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  11. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    SciTech Connect (OSTI)

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates that the proposed solutions to the investigated operating cycle length barriers are both feasible and consistent with sound design practice.

  12. Spent fuel sabotage aerosol test program :FY 2005-06 testing and aerosol data summary.

    SciTech Connect (OSTI)

    Gregson, Michael Warren; Brockmann, John E.; Nolte, O. (Fraunhofer institut fur toxikologie und experimentelle Medizin, Germany); Loiseau, O. (Institut de radioprotection et de Surete Nucleaire, France); Koch, W. (Fraunhofer institut fur toxikologie und experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno (Institut de radioprotection et de Surete Nucleaire, France); Pretzsch, Gunter Guido (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Billone, M. C. (Argonne National Laboratory, USA); Lucero, Daniel A.; Burtseva, T. (Argonne National Laboratory, USA); Brucher, W (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2006-10-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides source-term data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This document focuses on an updated description of the test program and test components for all work and plans made, or revised, primarily during FY 2005 and about the first two-thirds of FY 2006. It also serves as a program status report as of the end of May 2006. We provide details on the significant findings on aerosol results and observations from the recently completed Phase 2 surrogate material tests using cerium oxide ceramic pellets in test rodlets plus non-radioactive fission product dopants. Results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; status on determination of the spent fuel ratio, SFR (the ratio of respirable particles from real spent fuel/respirables from surrogate spent fuel, measured under closely matched test conditions, in a contained test chamber); and, measurements of enhanced volatile fission product species sorption onto respirable particles. We discuss progress and results for the first three, recently performed Phase 3 tests using depleted uranium oxide, DUO{sub 2}, test rodlets. We will also review the status of preparations and the final Phase 4 tests in this program, using short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. These data plus testing results and design are tailored to support and guide, follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage--aerosol test program, performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission, had significant inputs from, and is strongly supported and coordinated by both the U.S. and international program participants in Germany, France, and the U.K., as part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC.

  13. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Radiation Transport Methods (RTM) Delivers next-generation radiation transport tools to the virtual Reactor RTM Vision Statement Objectives and Strategies Next generation,...

  14. HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

    DOE Patents [OSTI]

    Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

    1960-06-14

    A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

  15. Standard Contract Amendment for New Reactors | Department of Energy

    Energy Savers [EERE]

    Amendment for New Reactors Standard Contract Amendment for New Reactors The following document is a contract amendment to a standard contract between the United States of America, represented by the U.S. Department of Energy, and a corporation organized and existing under the laws of a given state to dispose of Spent Nuclear Fuel (SNF)and/or High-Level Radioactive Waste (HRW). PDF icon New_Standard_Contract_Amendment.pdf More Documents & Publications Standard Contract for Disposal of SNF

  16. Operation of N Reactor and Fuels Fabrication Facilities, Hanford Reservation, Richland, Benton County, Washington: Environmental assessment

    SciTech Connect (OSTI)

    Not Available

    1980-08-01

    Environmental data, calculations and analyses show no significant adverse radiological or nonradiological impacts from current or projected future operations resulting from N Reactor, Fuels Fabrication and Spent Fuel Storage Facilities. Nonoccupational radiation exposures resulting from 1978 N Reactor operations are summarized and compared to allowable exposure limits.

  17. LWR spent fuel reduction by the removal of U and the compact storage of Pu with FP for long-term nuclear sustainability

    SciTech Connect (OSTI)

    Fukasawa, T.; Hoshino, K.; Takano, M.; Sato, S.; Shimazu, Y.

    2013-07-01

    Fast breeder reactors (FBR) nuclear fuel cycle is needed for long-term nuclear sustainability while preventing global warming and maximum utilizing the limited uranium (U) resources. The 'Framework for Nuclear Energy Policy' by the Japanese government on October 2005 stated that commercial FBR deployment will start around 2050 under its suitable conditions by the successive replacement of light water reactors (LWR) to FBR. Even after Fukushima Daiichi Nuclear Power Plant accident which made Japanese tendency slow down the nuclear power generation activities, Japan should have various options for energy resources including nuclear, and also consider the delay of FBR deployment and increase of LWR spent fuel (LWR-SF) storage amounts. As plutonium (Pu) for FBR deployment will be supplied from LWR-SF reprocessing and Japan will not possess surplus Pu, the authors have developed the flexible fuel cycle initiative (FFCI) for the transition from LWR to FBR. The FFCI system is based on the possibility to stored recycled materials (U, Pu)temporarily for a suitable period according to the FBR deployment rate to control the Pu demand/supply balance. This FFCI system is also effective after the Fukushima accident for the reduction of LWR-SF and future LWR-to-FBR transition. (authors)

  18. Microstructural characteristics of PWR spent fuel relative to its leaching behavior

    SciTech Connect (OSTI)

    Wilson, C.N.

    1985-11-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

  19. Maine Yankee: Making the Transition from an Operating Plant to an Independent Spent Fuel Storage Installation (ISFSI)

    SciTech Connect (OSTI)

    Norton, W.; McGough, M. S.

    2002-02-26

    The purpose of this paper is to describe the challenges faced by Maine Yankee Atomic Power Company in making the transition from an operating nuclear power plant to an Independent Spent Fuel Storage Installation (ISFSI). Maine Yankee (MY) is a 900-megawatt Combustion Engineering pressurized water reactor whose architect engineer was Stone & Webster. Maine Yankee was put into commercial operation on December 28, 1972. It is located on an 820-acre site, on the shores of the Back River in Wiscasset, Maine about 40 miles northeast of Portland, Maine. During its operating life, it generated about 1.2 billion kilowatts of power, providing 25% of Maine's electric power needs and serving additional customers in New England. Maine Yankee's lifetime capacity factor was about 67% and it employed more than 450 people. The decision was made to shutdown Maine Yankee in August of 1997, based on economic reasons. Once this decision was made planning began on how to accomplish safe and cost effective decommissioning of the plant by 2004 while being responsive to the community and employees.

  20. Spent Fuel Test-Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Final report

    SciTech Connect (OSTI)

    Patrick, W.C.

    1986-03-30

    In the Climax stock granite on the Nevada Test Site, eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized. When test data indicated that the test objectives were met during the 3-year storage phase, the spent-fuel canisters were retrieved and the thermal sources were de-energized. The project demonstrated the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner. In addition to emplacement and retrieval operations, three exchanges of spent-fuel assemblies between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. The test led to development of a technical measurements program. To meet these objectives, nearly 1000 instruments and a computer-based data acquisition system were deployed. Geotechnical, seismological, and test status data were recorded on a continuing basis for the three-year storage phase and six-month monitored cool-down of the test. This report summarizes the engineering and scientific endeavors which led to successful design and execution of the test. The design, fabrication, and construction of all facilities and handling systems are discussed, in the context of test objectives and a safety assessment. The discussion progresses from site characterization and experiment design through data acquisition and analysis of test data in the context of design calculations. 117 refs., 52 figs., 81 tabs.

  1. Thermal Hydraulic Analysis of Spent Fuel Casks

    Energy Science and Technology Software Center (OSTI)

    1997-10-08

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codesmore » for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.« less

  2. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, Daniel L. (Princeton, NJ)

    1988-01-01

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

  3. EM Completes Project to Maintain Water Quality of Spent Nuclear...

    Energy Savers [EERE]

    Completes Project to Maintain Water Quality of Spent Nuclear Fuel Basin at Idaho Site EM Completes Project to Maintain Water Quality of Spent Nuclear Fuel Basin at Idaho Site ...

  4. EM Safely and Efficiently Manages Spent Nuclear Fuel | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    (MVDS) is a dry storage system used at Idaho Engineering Laboratory and at Fort St. Vrain to store spent nuclear fuel. Each storage location holds one spent nuclear container. ...

  5. Scram signal generator

    DOE Patents [OSTI]

    Johanson, Edward W. (New Lenox, IL); Simms, Richard (Westmont, IL)

    1981-01-01

    A scram signal generating circuit for nuclear reactor installations monitors a flow signal representing the flow rate of the liquid sodium coolant which is circulated through the reactor, and initiates reactor shutdown for a rapid variation in the flow signal, indicative of fuel motion. The scram signal generating circuit includes a long-term drift compensation circuit which processes the flow signal and generates an output signal representing the flow rate of the coolant. The output signal remains substantially unchanged for small variations in the flow signal, attributable to long term drift in the flow rate, but a rapid change in the flow signal, indicative of a fast flow variation, causes a corresponding change in the output signal. A comparator circuit compares the output signal with a reference signal, representing a given percentage of the steady state flow rate of the coolant, and generates a scram signal to initiate reactor shutdown when the output signal equals the reference signal.

  6. Disposal options for burner ash from spent graphite fuel. Final study report November 1993

    SciTech Connect (OSTI)

    Pinto, A.P.

    1994-08-01

    Three major disposal alternatives are being considered for Fort St. Vrain Reactor (FSVR) and Peach Bottom Reactor (PBR) spent fuels: direct disposal of packaged, intact spent fuel elements; (2) removal of compacts to separate fuel into high-level waste (HLW) and low-level waste (LLW); and (3) physical/chemical processing to reduce waste volumes and produce stable waste forms. For the third alternative, combustion of fuel matrix graphite and fuel particle carbon coatings is a preferred technique for head-end processing as well as for volume reduction and chemical pretreatment prior to final fixation, packaging, and disposal of radioactive residuals (fissile and fertile materials together with fission and activation products) in a final repository. This report presents the results of a scoping study of alternate means for processing and/or disposal of fissile-bearing particles and ash remaining after combustion of FSVR and PBR spent graphite fuels. Candidate spent fuel ash (SFA) waste forms in decreasing order of estimated technical feasibility include glass-ceramics (GCs), polycrystalline ceramic assemblages (PCAs), and homogeneous amorphous glass. Candidate SFA waste form production processes in increasing order of estimated effort and cost for implementation are: low-density GCs via fuel grinding and simultaneous combustion and waste form production in a slagging cyclone combustor (SCC); glass or low-density GCs via fluidized bed SFA production followed by conventional melting of SFA and frit; PCAs via fluidized bed SFA production followed by hot isostatic pressing (HIPing) of SFA/frit mixtures; and high-density GCs via fluidized bed SFA production followed by HIPing of Calcine/Frit/SFA mixtures.

  7. January 28, 2016 Webinar - Borehole Disposal of Spent Radioactive Sources |

    Energy Savers [EERE]

    Department of Energy January 28, 2016 Webinar - Borehole Disposal of Spent Radioactive Sources January 28, 2016 Webinar - Borehole Disposal of Spent Radioactive Sources Performance & RIsk Assessment (P&RA) Community of Practice (CoP) Webinar - January 28, 2016 - Borehole Disposal of Spent Radioactive Sources (Dr. Matt Kozak, INTERA). Webinar Recording PDF icon Agenda & Webinar Instructions - January 28, 2016 - P&RA CoP Webinar PDF icon Borehole Disposal of Spent Sources

  8. Naval Spent Fuel Rail Shipment Accident Exercise Objectives

    Office of Environmental Management (EM)

    NAVAL SPENT FUEL RAIL SHIPMENT ACCIDENT EXERCISE OBJECTIVES * Familiarize stakeholders with the Naval spent fuel ACCIDENT EXERCISE OBJECTIVES Familiarize stakeholders with the Naval spent fuel shipping container characteristics and shipping practices * Gain understanding of how the NNPP escorts who accompany the spent fuel shipments will interact with civilian emergency services representatives g y p * Allow civilian emergency services agencies the opportunity to evaluate their response to a pp

  9. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    SciTech Connect (OSTI)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  10. Development of a thermal reclamation system for spent blasting abrasive

    SciTech Connect (OSTI)

    Bryan, B.B.; Mensinger, M.C.; Rehmat, A.G.

    1991-01-01

    Abrasive blasting is the most economical method for paint removal from large surface areas such as the hulls and tanks of oceangoing vessels. Tens of thousands of tons of spent abrasive are generated annually by blasting operations in private and US Navy shipyards. Some of this material is classified as hazardous waste, and nearly all of it is currently being either stockpiled or disposed in landfills. The rapid decline in available landfill space and corresponding rise in landfill tipping fees pose a severe problem for shipyard operators throughout the US. This paper discusses the results of a research and development program initiated by the Institute of Gas Technology and supported by the US Navy to develop and test a fluidized-bed thermal reclamation system for spent abrasive waste minimization. Bench- and pilot-scale reclaimer tests and reclaimed abrasive performance tests are described along with the current status of a program to build and test a 5-ton/hour prototype reclaimer at a US Navy shipyard.

  11. Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

    SciTech Connect (OSTI)

    Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven; Galic, Dom

    2004-08-01

    This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation--or neutrino oscillation--by measuring neutrinos produced in the fission reactions of a nuclear power plant. Diablo Canyon was selected as a candidate site because it allows the detectors to be situated underground in a tunnel close to the source of neutrinos (i.e., at a distance of several hundred meters from the nuclear power plant) while having suitable topography for shielding against cosmic rays. The detectors have to be located underground to minimize the cosmic ray-related background noise that can mimic the signal of reactor neutrino interactions in the detector. Three Pliocene-Miocene marine sedimentary units dominate the geology of Diablo Canyon: the Pismo Formation, the Monterey Formation, and the Obispo Formation. The area is tectonically active, located east of the active Hosgri Fault and in the southern limb of the northwest trending Pismo Syncline. Most of the potential tunnel for the neutrino detector lies within the Obispo Formation. Review of previous geologic studies, observations from a field visit, and selected physical and mechanical properties of rock samples collected from the site provided baseline geological information used in developing a preliminary estimate for tunneling construction cost. Gamma-ray spectrometric results indicate low levels of radioactivity for uranium, thorium, and potassium. Grain density, bulk density, and porosity values for these rock samples range from 2.37 to 2.86 g/cc, 1.41 to 2.57 g/cc, and 1.94 to 68.5% respectively. Point load, unconfined compressive strength, and ultrasonic velocity tests were conducted to determine rock mechanical and acoustic properties. The rock strength values range from 23 to 219 MPa and the Poisson's ratio from 0.1 to 0.38. Potential geologic hazards in the Diablo Canyon area were identified and described to provide an overall picture of processes that may affect tunnel construction activities.

  12. Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

    SciTech Connect (OSTI)

    Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven; Galic, Dom

    2004-06-11

    This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation or neutrino oscillation by measuring neutrinos produced in the fission reactions of a nuclear power plant. Diablo Canyon was selected as a candidate site because it allows the detectors to be situated underground in a tunnel close to the source of neutrinos (i.e., at a distance of several hundred meters from the nuclear power plant) while having suitable topography for shielding against cosmic rays. The detectors have to be located underground to minimize the cosmic ray-related background noise that can mimic the signal of reactor neutrino interactions in the detector. Three Pliocene-Miocene marine sedimentary units dominate the geology of Diablo Canyon: the Pismo Formation, the Monterey Formation, and the Obispo Formation. The area is tectonically active, located east of the active Hosgri Fault and in the southern limb of the northwest trending Pismo Syncline. Most of the potential tunnel for the neutrino detector lies within the Obispo Formation. Review of previous geologic studies, observations from a field visit, and selected physical and mechanical properties of rock samples collected from the site provided baseline geological information used in developing a preliminary estimate for tunneling construction cost. Gamma-ray spectrometric results indicate low levels of radioactivity for uranium, thorium, and potassium. Grain density, bulk density, and porosity values for these rock samples range from 2.37 to 2.86 g/cc, 1.41 to 2.57 g/cc, and 1.94 to 68.5 percent respectively. Point load, unconfined compressive strength, and ultrasonic velocity tests were conducted to determine rock mechanical and acoustic properties. The rock strength values range from 23 to 219 MPa and the Poisson's ratio from 0.1 to 0.38. Potential geologic hazards in the Diablo Canyon area were identified and described to provide an overall picture of processes that may affect tunnel construction activities.

  13. Naval Spent Fuel Rail Shipment Accident Exercise Objectives | Department of

    Office of Environmental Management (EM)

    Energy Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives PDF icon Naval Spent Fuel Rail Shipment Accident Exercise Objectives More Documents & Publications TEC Meeting Summaries - April 2005 Presentations TEC Meeting Summaries - January - February 2007 Presentations NTSF 2014 Meeting Agenda

  14. Spent fuel dissolution rates as a function of burnup and water chemistry

    SciTech Connect (OSTI)

    Gray, W.J.

    1998-06-01

    To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of {sup 129}I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and {approximately} 65 MWd/kgM. (2) Oxidation of spent fuel up to the U{sub 4}O{sub 9+x} stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of {sup 129}I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and {sup 129}I gap inventory for US LWR fuels.

  15. Solid tags for identifying failed reactor components

    DOE Patents [OSTI]

    Bunch, Wilbur L. (Richland, WA); Schenter, Robert E. (Richland, WA)

    1987-01-01

    A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.

  16. Measurement of Spent Fuel Assemblies - Overview of the Status of the Technology for Initiating Discussion at NATIONAL RESEARCH CENTRE KURCHATOV INSTITUTE June 2013

    SciTech Connect (OSTI)

    SISKIND B.; N /A

    2013-06-03

    This presentation provides an overview of the status of the technology for the measurement of the fissile material content of spent nuclear reactor fuel. The emphasis is on how the needs of the U.S. Nuclear Regulatory Commission and the International Atomic Energy Agency are met by the available technology and what more needs to be done in this area.

  17. The EU Approach for Responsible and Safe Management of Spent Fuel and Radioactive Waste - 12118

    SciTech Connect (OSTI)

    Blohm-Hieber, Ute; Necheva, Christina [European Commission, Directorate-General for Energy, Luxembourg L-2920 (Luxembourg)

    2012-07-01

    In July 2011 legislation on responsible and safe management of spent fuel and radioactive waste was adopted in the European Union (EU). It aims at ensuring a high level of safety, avoiding undue burdens on future generations and enhancing transparency. EU Member States are responsible for the management of their spent fuel and/or radioactive waste. Each Member State remains free to define its fuel cycle policy. The spent fuel can be regarded either as a valuable resource that may be reprocessed or as radioactive waste that is destined for direct disposal. Whatever option is chosen, the disposal of high level waste, separated at reprocessing, or of spent fuel regarded as waste should be considered. The storage of radioactive waste, including long-term storage, is an interim solution, but not an alternative to disposal. To this end, each Member State has to establish, maintain and implement national policy, framework and programme for management of spent fuel and/or radioactive waste in the long term. Member States will invite international peer reviews to ensure that high safety standards are achieved. The EU approach is anchored in internationally endorsed principles and requirements of the IAEA safety standards and the Joint Convention and in this context makes them legally binding and enforceable in the EU. The EU approach of regulating the management of spent fuel and radioactive waste is anchored in the competence of the national regulatory authorities and in the internationally endorsed principles and requirements of the IAEA Safety Standards and the Joint Convention. Member States have to report to the Commission on the implementation of Directive 2011/70/Euratom for the first time by 23 August 2015, and every 3 years thereafter, taking advantage of the review and reporting under the Joint Convention. On the basis of the Member States' reports, the Commission will submit to the European Parliament and the Council a report on progress made and an inventory of radioactive waste and spent fuel present in the EU territory and the future prospects. Directive 2011/70/Euratom is a logical next step after the Council Directive 2009/71/Euratom on the nuclear safety of nuclear installations. The EU is the first major regional actor providing a binding legal framework on nuclear safety and on responsible and safe management of spent fuel and radioactive waste, and thus is a real model to progress spent fuel and waste management in a safe and responsible manner. (authors)

  18. Initial Requirements for Gas-Cooled Fast Reactor (GFR) System Design, Performance, and Safety Analysis Models

    SciTech Connect (OSTI)

    Kevan D. Weaver; Thomas Y. C. Wei

    2004-08-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  19. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  20. Reactor refueling machine simulator

    SciTech Connect (OSTI)

    Rohosky, T.L.; Swidwa, K.J.

    1987-10-13

    This patent describes in combination: a nuclear reactor; a refueling machine having a bridge, trolley and hoist each driven by a separate motor having feedback means for generating a feedback signal indicative of movement thereof. The motors are operable to position the refueling machine over the nuclear reactor for refueling the same. The refueling machine also has a removable control console including means for selectively generating separate motor signals for operating the bridge, trolley and hoist motors and for processing the feedback signals to generate an indication of the positions thereof, separate output leads connecting each of the motor signals to the respective refueling machine motor, and separate input leads for connecting each of the feedback means to the console; and a portable simulator unit comprising: a single simulator motor; a single simulator feedback signal generator connected to the simulator motor for generating a simulator feedback signal in response to operation of the simulator motor; means for selectively connecting the output leads of the console to the simulator unit in place of the refueling machine motors, and for connecting the console input leads to the simulator unit in place of the refueling machine motor feedback means; and means for driving the single simulator motor in response to any of the bridge, trolley or hoist motor signals generated by the console and means for applying the simulator feedback signal to the console input lead associated with the motor signal being generated by the control console.

  1. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R.

    1996-07-01

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  2. Radioactive Waste Management: Study of Spent Fuel Dissolution Rates in Geological Storage Using Dosimetry Modeling and Experimental Verification

    SciTech Connect (OSTI)

    Hansen, Brady; Miller, William

    2011-10-28

    This research will provide improved predictions into the mechanisms and effects of radiolysis on spent nuclear fuel dissolution in a geological respository through accurate dosimetry modeling of the dose to water, mechanistic chemistry modeling of the resulting radiolytic reactions and confirmatory experimental measurements. This work will combine effort by the Nuclear Science and Engineering Institute (NSEI) and the Missouri University Research Reactor (MURR) at the University of Missouri-Columbia, and the expertise and facilities at the Pacific Northwest National Laboratory (PNNL).

  3. Spent Sealed Sources Management in Switzerland - 12011

    SciTech Connect (OSTI)

    Beer, H.F.

    2012-07-01

    Information is provided about the international recommendations for the safe management of disused and spent sealed radioactive sources wherein the return to the supplier or manufacturer is encouraged for large radioactive sources. The legal situation in Switzerland is described mentioning the demand of minimization of radioactive waste as well as the situation with respect to the interim storage facility at the Paul Scherrer Institute (PSI). Based on this information and on the market situation with a shortage of some medical radionuclides the management of spent sealed sources is provided. The sources are sorted according to their activity in relation to the nuclide-specific A2-value and either recycled as in the case of high active sources or conditioned as in the case for sources with lower activity. The results are presented as comparison between recycled and conditioned activity for three selected nuclides, i.e. Cs-137, Co-60 and Am-241. (author)

  4. Spent fuel container alignment device and method

    DOE Patents [OSTI]

    Jones, Stewart D.; Chapek, George V.

    1996-01-01

    An alignment device is used with a spent fuel shipping container including a plurality of fuel pockets for spent fuel arranged in an annular array and having a rotatable cover including an access opening therein. The alignment device includes a lightweight plate which is installed over the access opening of the cover. A laser device is mounted on the plate so as to emit a laser beam through a laser admittance window in the cover into the container in the direction of a pre-established target associated with a particular fuel pocket. An indexing arrangement on the container provides an indication of the angular position of the rotatable cover when the laser beam produced by the laser is brought into alignment with the target of the associated fuel pocket.

  5. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

    SciTech Connect (OSTI)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier; Klennert, Lindsay A.; Nolte, Oliver; Molecke, Martin Alan; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Brucher, Wenzel; Steyskal, Michele D.

    2008-03-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively supported and coordinated by both the U.S. and international program participants in Germany, France, and others, as part of the International Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC).

  6. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.

    1960-11-22

    A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

  7. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  8. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul (Pittsburgh, PA)

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  9. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  10. Spent fuel pool analysis using TRACE code

    SciTech Connect (OSTI)

    Sanchez-Saez, F.; Carlos, S.; Villanueva, J. F.; Martorell, S.

    2012-07-01

    The storage requirements of Spent Fuel Pools have been analyzed with the purpose to increase their rack capacities. In the past, the thermal limits have been mainly evaluated with conservative codes developed for this purpose, although some works can be found in which a best estimate code is used. The use of best estimate codes is interesting as they provide more realistic calculations and they have the capability of analyzing a wide range of transients that could affect the Spent Fuel Pool. Two of the most representative thermal-hydraulic codes are RELAP-5 and TRAC. Nowadays, TRACE code is being developed to make use of the more favorable characteristics of RELAP-5 and TRAC codes. Among the components coded in TRACE that can be used to construct the model, it is interesting to use the VESSEL component, which has the capacity of reproducing three dimensional phenomena. In this work, a thermal-hydraulic model of the Maine Yankee spent fuel pool using the TRACE code is developed. Such model has been used to perform a licensing calculation and the results obtained have been compared with experimental measurements made at the pool, showing a good agreement between the calculations predicted by TRACE and the experimental data. (authors)

  11. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  12. N-Reactor (U-metal) Fuel Characteristics for Disposal Criticality Analysis

    SciTech Connect (OSTI)

    Taylor, Larry Lorin

    2000-05-01

    DOE-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into nine characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. Additionally, the criticality analysis will also require data to support design of the canister internals, thermal, and radiation shielding. The purpose of this report is to consolidate and provide in a concise format, material and information/data needed to perform supporting analyses to qualify N-Reactor fuels for acceptance into the designated repository. The N Reactor fuels incorporate zirconium cladding and uranium metal with unique fabrication details in terms of physical size, and method of construction. The fuel construction and post-irradiation handling have created attendant issues relative to cladding failure in the underwater storage environment. These fuels were comprised of low-enriched metal (0.947 to 1.25 wt% 235U) that were originally intended to generate weapons-grade plutonium for national defense. Modifications in subsequent fuel design and changes in the mode of reactor operation in later years were focused more toward power production.

  13. Self isolating high frequency saturable reactor

    DOE Patents [OSTI]

    Moore, James A. (Powell, TN)

    1998-06-23

    The present invention discloses a saturable reactor and a method for decoupling the interwinding capacitance from the frequency limitations of the reactor so that the equivalent electrical circuit of the saturable reactor comprises a variable inductor. The saturable reactor comprises a plurality of physically symmetrical magnetic cores with closed loop magnetic paths and a novel method of wiring a control winding and a RF winding. The present invention additionally discloses a matching network and method for matching the impedances of a RF generator to a load. The matching network comprises a matching transformer and a saturable reactor.

  14. Numerical approach for the voloxidation process of an advanced spent fuel conditioning process (ACP)

    SciTech Connect (OSTI)

    Park, Byung Heung; Jeong, Sang Mun; Seo, Chung-Seok

    2007-07-01

    A voloxidation process is adopted as the first step of an advanced spent fuel conditioning process in order to prepare the SF oxide to be reduced in the following electrolytic reduction process. A semi-batch type voloxidizer was devised to transform a SF pellet into powder. In this work, a simple reactor model was developed for the purpose of correlating a gas phase flow rate with an operation time as a numerical approach. With an assumption that a solid phase and a gas phase are homogeneous in a reactor, a reaction rate for an oxidation was introduced into a mass balance equation. The developed equation can describe a change of an outlet's oxygen concentration including such a case that a gas flow is not sufficient enough to continue a reaction at its maximum reaction rate. (authors)

  15. Control of reactor coolant flow path during reactor decay heat removal

    DOE Patents [OSTI]

    Hunsbedt, Anstein N. (Los Gatos, CA)

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  16. American National Standard: design criteria for an independent spent-fuel-storage installation (water pool type)

    SciTech Connect (OSTI)

    Not Available

    1981-01-01

    This standard provides design criteria for systems and equipment of a facility for the receipt and storage of spent fuel from light water reactors. It contains requirements for the design of major buildings and structures including the shipping cask unloading and spent fuel storage pools, cask decontamination, unloading and loading areas, and the surrounding buildings which contain radwaste treatment, heating, ventilation and air conditioning, and other auxiliary systems. It contains requirements and recommendations for spent fuel storage racks, special equipment and area layout configurations, the pool structure and its integrity, pool water cleanup, ventilation, residual heat removal, radiation monitoring, fuel handling equipment, cask handling equipment, prevention of criticality, radwaste control and monitoring systems, quality assurance requirements, materials accountability, and physical security. Such an installation may be independent of both a nuclear power station and a reprocessing facility or located adjacent to any of these facilities in order to share selected support systems. Support systems shall not include a direct means of transferring fuel assemblies from the nuclear facility to the installation.

  17. Spent Fuel Test - Climax: technical measurements. Interim report, fiscal year 1982

    SciTech Connect (OSTI)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.; Carlson, R.C.; Durham, W.B.; Hage, G.L.; Majer, E.L.; Montan, D.N.; Nyholm, R.A.; Rector, N.L.

    1983-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted for the US Department of Energy (DOE) under the technical direction of the Lawrence Livermore National Laboratory (LLNL). Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April to May 1980, thus initiating a test with a planned 3- to 5-year fuel storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Three exchanges of spent fuel between the SFT-C and a surface storage facility furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and two previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 2-1/2 years of the test on more than 900 channels. Data continue to be acquired from the test. Some data are now available for analysis and are presented here. Highlights of activities this year include analysis of fracture data obtained during site characterization, laboratory studies of radiation effects and drilling damage in Climax granite, improved calculations of near-field heat transfer and thermomechanical response, a ventilation effects study, and further development of the data acquisition and management systems.

  18. Characterization program management plan for Hanford K Basin Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Lawrence, L.A.

    1995-10-18

    A management plan was developed for Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratories (PNL) to work together on a program to provide characterization data to support removal, conditioning and subsequent dry storage of the spent nuclear fuels stored at the Hanford K Basins. The Program initially supports gathering data to establish the current state of the fuel in the two basins. Data Collected during this initial effort will apply to all SNF Project objectives. N Reactor fuel has been degrading with extended storage resulting in release of material to the basin water in K East and to the closed conisters in K West. Characterization of the condition of these materials and their responses to various conditioning processes and dry storage environments are necessary to support disposition decisions. Characterization will utilize the expertise and capabilities of WHC and PNL organizations to support the Spent Nuclear Fuels Project goals and objectives. This Management Plan defines the structure and establishes the roles for the participants providing the framework for WHC and PNL to support the Spent Nuclear Fuels Project at Hanford

  19. Measurement of plutonium in spent nuclear fuel by self-induced x-ray fluorescence

    SciTech Connect (OSTI)

    Hoover, Andrew S; Rudy, Cliff R; Tobin, Steve J; Charlton, William S; Stafford, A; Strohmeyer, D; Saavadra, S

    2009-01-01

    Direct measurement of the plutonium content in spent nuclear fuel is a challenging problem in non-destructive assay. The very high gamma-ray flux from fission product isotopes overwhelms the weaker gamma-ray emissions from plutonium and uranium, making passive gamma-ray measurements impossible. However, the intense fission product radiation is effective at exciting plutonium and uranium atoms, resulting in subsequent fluorescence X-ray emission. K-shell X-rays in the 100 keV energy range can escape the fuel and cladding, providing a direct signal from uranium and plutonium that can be measured with a standard germanium detector. The measured plutonium to uranium elemental ratio can be used to compute the plutonium content of the fuel. The technique can potentially provide a passive, non-destructive assay tool for determining plutonium content in spent fuel. In this paper, we discuss recent non-destructive measurements of plutonium X-ray fluorescence (XRF) signatures from pressurized water reactor spent fuel rods. We also discuss how emerging new technologies, like very high energy resolution microcalorimeter detectors, might be applied to XRF measurements.

  20. The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life

    SciTech Connect (OSTI)

    J. H. Wolfram; R. E. Mizia; R. Jex; L. Nelson; K. M. Garcia

    1996-10-01

    A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination.

  1. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    1995-09-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  2. THE BIRTH OF NUCLEAR-GENERATED ELECTRICITY

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    BIRTH OF NUCLEAR-GENERATED ELECTRICITY The first time that electricity was generated from nuclear energy occurred in an experimental breeder reactor in Idaho in 1951. The idea for a breeder reactor (a reactor that could produce more fuel than it uses) first occurred to scientists working on the nation's wartime atomic energy program in the early 1940's. Experimental evidence indicated that the breeding of nuclear fuel was possible in a properly designed reactor, but time and resources were not

  3. Development of Materials for Supercritical-Water-Cooled Reactor

    Broader source: Energy.gov [DOE]

    Supercritical-Water-Cooled Reactor (SCWR) was selected as one of the promising candidates in Generation IV reactors for its prominent advantages; those are the high thermal efficiency, the system...

  4. REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

    DOE Patents [OSTI]

    Loeb, E.

    1961-01-17

    A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

  5. Geologic repository design and disposal: GNEP spent fuel processing-waste volume

    SciTech Connect (OSTI)

    Bauer, T.H.; Wigeland, R.A.

    2007-07-01

    Previous work has shown that removal of key heat generating elements from spent fuel would allow greater utilization of space in a geologic repository such as Yucca Mountain by factors of 100 or more without increasing the estimated peak dose rate to an exposed individual. However, achieving such utilization increases within a repository storage drift requires the density of the remaining fission products, actinide elements, etc. to be increased by roughly the same factor as the utilization increase, itself. This paper analyzes several alternative drift configurations possible within a designated repository area that could: (1) allow greater volume for waste storage and (2) maintain significant utilization benefit. For a representative range of GNEP-generated waste streams, computed results show that increase in repository area space utilization by a factor {approx}100 can be maintained with such configurations as long as waste stream volume can be reduced from that of the original spent fuel by a factor of {approx}10. (authors)

  6. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  7. Nuclear reactor characteristics and operational history

    Gasoline and Diesel Fuel Update (EIA)

    1. Capacity and Generation, Table 3. Characteristics and Operational History Table 2. U.S. Nuclear Reactor Ownership Data PDF XLS Plant/Reactor Name Generator ID Utility Name - Operator Owner Name % Owned Arkansas Nuclear One 1 Entergy Arkansas Inc Entergy Arkansas Inc 100 Arkansas Nuclear One 2 Entergy Arkansas Inc Entergy Arkansas Inc 100 Beaver Valley 1 FirstEnergy Nuclear Operating Company FirstEnergy Nuclear Generation Corp 100 Beaver Valley 2 FirstEnergy Nuclear Operating Company

  8. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    SciTech Connect (OSTI)

    F. Habashi

    1998-06-26

    The DOE Spent Nuclear Fuel Disposal Container (SNF DC) supports the confinement and isolation of waste within the Engineered Barrier System of the Mined Geologic Disposal System (MGDS). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the access mains, and emplaced in emplacement drifts. The DOE Spent Nuclear Fuel Disposal Container provides long term confinement of DOE SNF waste, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The DOE SNF Disposal Containers provide containment of waste for a designated period of time, and limit radionuclide release thereafter. The disposal containers maintain the waste in a designated configuration, withstand maximum handling and rockfall loads, limit the individual waste canister temperatures after emplacement. The disposal containers also limit the introduction of moderator into the disposal container during the criticality control period, resist corrosion in the expected repository environment, and provide complete or limited containment of waste in the event of an accident. Multiple disposal container designs may be needed to accommodate the expected range of DOE Spent Nuclear Fuel. The disposal container will include outer and inner barrier walls and outer and inner barrier lids. Exterior labels will identify the disposal container and contents. Differing metal barriers will support the design philosophy of defense in depth. The use of materials with different failure mechanisms prevents a single mode failure from breaching the waste package. The corrosion-resistant inner barrier and inner barrier lid will be constructed of a high-nickel alloy and the corrosion-allowance outer barrier and outer barrier lid will be made of carbon steel. The DOE Spent Nuclear Fuel Disposal Containers interface with the emplacement drift environment by transferring heat from the waste to the external environment and by protecting the DOE waste canisters and their contents from damage/degradation by the external environment. The disposal containers also interface with the SNF by limiting access of moderator and oxidizing agents to the waste. The disposal containers interface with the Ex-Container System's emplacement drift disposal container supports. The disposal containers interface with the Canister Transfer System, Waste Emplacement System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and remediation of the disposal container.

  9. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom; Howard, Rob L

    2014-01-01

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

  10. The effect of the composition of plutonium loaded on the reactivity change and the isotopic composition of fuel produced in a fast reactor

    SciTech Connect (OSTI)

    Blandinskiy, V. Yu.

    2014-12-15

    This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.

  11. Dry-vault storage of spent fuel at the CASCAD facility

    SciTech Connect (OSTI)

    Baillif, L.; Guay, M.

    1989-01-01

    A new modular dry storage vault concept using vertical metallic wells cooled by natural convection has been developed by the Commissariat a l'Energie Atomique and Societe Generale pour les Techniques Nouvelles to accommodate special fuels for high-level wastes. Basic specifications and design criteria have been followed to guarantee a double containment system and cooling to maintain the fuel below an acceptable temperature. The double containment is provided by two static barriers: At the reactor, fuels are placed in containers playing the role of the first barrier; the storage wells constitute the second barrier. Spent fuel placed in wells is cooled by natural convection: a boundary layer is created along the outer side of the well. The heated air rises along the well leading to a thermosiphon flow that extracts the heat released. For heat transfer, studies, computations, and experimental tests have been carried out to calculate and determine the temperature of the containers and the fuel rod temperatures in various situations. The CASCAD vault storage can be applied to light water reactor (LWR) fuels without any difficulties if two requirements are satisfied: (1) Spend fuels have to be inserted in tight canisters. (2) Spent fuels have to be received only after a minimum decay time of 5 yr.

  12. Public information circular for shipments of irradiated reactor fuel. Revision 9

    SciTech Connect (OSTI)

    Not Available

    1993-03-01

    This circular has been prepared to provide information on the shipment of irradiated reactor fuel (spent fuel) subject to regulation by the Nuclear Regulatory Commission (NRC), and to meet the requirements of Public Law 96--295. The report provides a brief description of NRC authority for certain aspects of transporting spent fuel. It provides descriptive statistics on spent fuel shipments regulated by the NRC from 1979 to 1992. It also lists detailed highway and railway segments used within each state from October 1, 1987 through December 31, 1992.

  13. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    SciTech Connect (OSTI)

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  14. Spent fuel management fee methodology and computer code user's manual.

    SciTech Connect (OSTI)

    Engel, R.L.; White, M.K.

    1982-01-01

    The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively.

  15. Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    B.R. Westphal; D. Vaden; S.X. Li; G.L. Fredrickson; R.D. Mariani

    2009-09-01

    During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontally displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor metallic fuel will be shown to be of no consequence to reactor performance.

  16. HOW MANY DID YOU SAY? HISTORICAL AND PROJECTED SPENT NUCLEAR FUEL SHIPMENTS IN THE UNITED STATES, 1964 - 2048

    SciTech Connect (OSTI)

    Halstead, Robert J.; Dilger, Fred

    2003-02-27

    No comprehensive, up-to-date, official database exists for spent nuclear fuel shipments in the United States. The authors review the available data sources, and conclude that the absence of such a database can only be rectified by a major research effort, similar to that carried out by Oak Ridge National Laboratory (ORNL) in the early 1990s. Based on a variety of published references, and unpublished data from the U.S. Nuclear Regulatory Commission (NRC), the authors estimate cumulative U.S. shipments of commercial spent fuel for the period 1964-2001. The cumulative estimates include quantity shipped, number of cask-shipments, and shipment-miles, by truck and by rail. The authors review previous estimates of future spent fuel shipments, including contractor reports prepared for the U.S. Department of Energy (DOE), NRC, and the State of Nevada. The DOE Final Environmental Impact Statement (FEIS) for Yucca Mountain includes projections of spent nuclear fuel and high-level radioactive was te shipments for two inventory disposal scenarios (24 years and 38 years) and two national transportation modal scenarios (''mostly legal-weight truck'' and ''mostly rail''). Commercial spent fuel would compromise about 90 percent of the wastes shipped to the repository. The authors estimate potential shipments to Yucca Mountain over 38 years (2010-2048) for the DOE ''mostly legal-weight truck'' and ''mostly rail'' scenarios, and for an alternative modal mix scenario based on current shipping capabilities of the 72 commercial reactor sites. The cumulative estimates of future spent fuel shipments include quantity shipped, number of cask-shipments, and shipment-miles, by legal-weight truck, heavy-haul truck, rail and barge.

  17. Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Lessons Learned Report, NNSA, Feb 2010 | Department of Energy Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy Lessons Learned Report, NNSA, Feb 2010 Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy Lessons Learned Report, NNSA, Feb 2010 PDF icon 000466 Idaho Spent Fuel Facility (ISFF) Project Appropriate Acquisition Strategy Lessons Learned Report Feb 2011.pdf More Documents & Publications Integrated Biorefinery Research Facility (IBRF

  18. EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing

    Energy Savers [EERE]

    U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany | Department of Energy 77: Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany SUMMARY This EA will evaluate the potential environmental impacts of a DOE proposal to accept spent nuclear fuel from the

  19. EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina | Department

    Office of Environmental Management (EM)

    of Energy 79: Spent Nuclear Fuel Management, Aiken, South Carolina EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina SUMMARY The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. PUBLIC COMMENT

  20. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  1. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  2. Pyroprocess for processing spent nuclear fuel

    DOE Patents [OSTI]

    Miller, William E.; Tomczuk, Zygmunt

    2002-01-01

    This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF.sub.4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF.sub.4 with a eutectic melting point of 500.degree. C. Prior to lowering the basket, the salt is heated to a temperature of between 550.degree. C. and 700.degree. C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF.sub.6. In addition, after dissolution, the basket contains PuO.sub.2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

  3. Systems impacts of spent fuel disassembly alternatives

    SciTech Connect (OSTI)

    Not Available

    1984-07-01

    Three studies were completed to evaluate four alternatives to the disposal of intact spent fuel assemblies in a geologic repository. A preferred spent fuel waste form for disposal was recommended on consideration of (1) package design and fuel/package interaction, (2) long-term, in-repository performance of the waste form, and (3) overall process performance and costs for packaging, handling, and emplacement. The four basic alternative waste forms considered were (1) end fitting removal, (2) fission gas venting, (3) disassembly and close packing, and (4) shearing/immobilization. None of the findings ruled out any alternative on the basis of waste package considerations or long-term performance of the waste form. The third alternative offers flexibility in loading that may prove attractive in the various geologic media under consideration, greatly reduces the number of packages, and has the lowest unit cost. These studies were completed in October, 1981. Since then Westinghouse Electric Corporation and the Office of Nuclear Waste Isolation have completed studies in related fields. This report is now being published to provide publicly the background material that is contained within. 47 references, 28 figures, 31 tables.

  4. Evaluation of Options for Permanent Geologic Disposal of Spent...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    This study provides a technical basis for informing policy decisions regarding strategies for the management and permanent disposal of spent nuclear fuel (SNF) and high-level ...

  5. Management of Spent Dessicant from Vapour Recovery Dryers

    Office of Environmental Management (EM)

    *Significant literature exists on the performance characteristics and on tritium "memory" effects *Focus of request is more on the disposal of spent molecular sieves Purpose...

  6. Deep Borehole Disposal of Spent Fuel. Brady, Patrick V. Abstract...

    Office of Scientific and Technical Information (OSTI)

    Spent Fuel. Brady, Patrick V. Abstract not provided. Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States) USDOE National Nuclear Security Administration (NNSA)...

  7. Spent Nuclear Fuel project integrated safety management plan

    SciTech Connect (OSTI)

    Daschke, K.D.

    1996-09-17

    This document is being revised in its entirety and the document title is being revised to ``Spent Nuclear Fuel Project Integrated Safety Management Plan.

  8. Huizenga leads safety of spent fuel management, radioactive waste...

    National Nuclear Security Administration (NNSA)

    Huizenga leads safety of spent fuel management, radioactive waste management meeting in Vienna | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People...

  9. Plutonium Consumption Program, CANDU Reactor Project final report

    SciTech Connect (OSTI)

    Not Available

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  10. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, D.L.

    1987-09-04

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

  11. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  12. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    SciTech Connect (OSTI)

    2000-10-12

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in the emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Multiple boiling water reactor (BWR) and pressurized water reactor (PWR) disposal container designs are needed to accommodate the expected range of spent fuel assemblies and provide long-term confinement of the commercial SNF. The disposal container will include outer and inner cylinder walls, outer cylinder lids (two on the top, one on the bottom), inner cylinder lids (one on the top, one on the bottom), and an internal metallic basket structure. Exterior labels will provide a means by which to identify the disposal container and its contents. The two metal cylinders, in combination with the cladding, Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lid will be made of high-nickel alloy. The basket will assist criticality control, provide structural support, and improve heat transfer. The Uncanistered SNF Disposal Container System interfaces with the emplacement drift environment and internal waste by transferring heat from the SNF to the external environment and by protecting the SFN assemblies and their contents from damage/degradation by the external environment. The system also interfaces with the SFN by limiting access of moderator and oxidizing agents of the SFN. The waste package interfaces with the Emplacement Drift System's emplacement drift pallets upon which the wasted packages are placed. The disposal container interfaces with the Assembly Transfer System, Waste Emplacement/Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and retrieval of the disposal container/waste package.

  13. REACTOR MONITORING

    DOE Patents [OSTI]

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  14. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    SciTech Connect (OSTI)

    IRWIN, J.J.

    2000-11-18

    The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 100K Area fuel storage water basins (i.e., the K East and K West Basins) at the US. Department of Energy Hanford Site in Southeastern Washington state. Removal of free water is necessary to halt water-induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB). The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the 100K inner area using existing roadways. The CVDF will remove free. water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed, leak tested, and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.) The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drying process and transfer process water removed from the MCO back to the K Basins.

  15. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    SciTech Connect (OSTI)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

  16. EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering

    Office of Environmental Management (EM)

    Laboratory Environmental Restoration and Waste Management Programs | Department of Energy 3: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs SUMMARY This EIS considers programmatic (DOE-wide) alternative approaches to safely, efficiently, and responsibly manage existing and

  17. Electrochemical cell apparatus having axially distributed entry of a fuel-spent fuel mixture transverse to the cell lengths

    DOE Patents [OSTI]

    Reichner, Philip (Plum Borough, PA); Dollard, Walter J. (Churchill Borough, PA)

    1991-01-01

    An electrochemical apparatus (10) is made having a generator section (22) containing axially elongated electrochemical cells (16), a fresh gaseous feed fuel inlet (28), a gaseous feed oxidant inlet (30), and at least one gaseous spent fuel exit channel (46), where the spent fuel exit channel (46) passes from the generator chamber (22) to combine with the fresh feed fuel inlet (28) at a mixing apparatus (50), reformable fuel mixture channel (52) passes through the length of the generator chamber (22) and connects with the mixing apparatus (50), that channel containing entry ports (54) within the generator chamber (22), where the axis of the ports is transverse to the fuel electrode surfaces (18), where a catalytic reforming material is distributed near the reformable fuel mixture entry ports (54).

  18. Steam generator tube failures

    SciTech Connect (OSTI)

    MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

    1996-04-01

    A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service.

  19. Advancing Small Modular Reactors: How We're Supporting Next-Gen Nuclear Energy Technology

    Broader source: Energy.gov [DOE]

    Learn about the Energy Department's support for the next-generation nuclear energy technology -- small modular reactors.

  20. A method for determining the spent-fuel contribution to transport cask containment requirements

    SciTech Connect (OSTI)

    Sanders, T.L.; Seager, K.D.; Rashid, Y.R.; Barrett, P.R.; Malinauskas, A.P.; Einziger, R.E.; Jordan, H.; Duffey, T.A.; Sutherland, S.H.; Reardon, P.C.

    1992-11-01

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs.

  1. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Thermal Hydraulics Methods (THM) Delivers next-generation thermal-hydraulic simulation tools to Virtual Environment for Reactor Applications (VERA) Thermal Hydraulics Methods...

  2. HTGR Spent Fuel Treatment Program. HTGR Spent Fuel Treatment Development Program Plan

    SciTech Connect (OSTI)

    Not Available

    1984-12-01

    The spent fuel treatment (SFT) program plan addresses spent fuel volume reduction, packaging, storage, transportation, fuel recovery, and disposal to meet the needs of the HTGR Lead Plant and follow-on plants. In the near term, fuel refabrication will be addressed by following developments in fresh fuel fabrication and will be developed in the long term as decisions on the alternatives dictate. The formulation of this revised program plan considered the implications of the Nuclear Waste Policy Act of 1982 (NWPA) which, for the first time, established a definitive national policy for management and disposal of nuclear wastes. Although the primary intent of the program is to address technical issues, the divergence between commercial and government interests, which arises as a result of certain provisions of the NWPA, must be addressed in the economic assessment of technically feasible alternative paths in the management of spent HTGR fuel and waste. This new SFT program plan also incorporates a significant cooperative research and development program between the United States and the Federal Republic of Germany. The major objective of this international program is to reduce costs by avoiding duplicate efforts.

  3. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    SciTech Connect (OSTI)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  4. Selective Separation of Cs and Sr from LiCl-Based Salt for Electrochemical Processing of Oxide Spent Nuclear Fuel

    SciTech Connect (OSTI)

    P Sachdev

    2008-07-01

    Electrochemical processing technology is currently being used for the treatment of metallic spent fuel from the Experimental Breeder Reactor-II at Idaho National Laboratory. The treatment of oxide-based spent nuclear fuel via electrochemical processing is possible provided there is a front-end oxide reduction step. During this reduction process, certain fission products, including Cs and Sr, partition into the salt phase and form chlorides. Both solid state and molten LiCl-zeolite-A ion exchange tests were conducted for selectively removing Cs and Sr from LiCl-based salt. The solid-state tests produced in excess of 99% removal of Cs and Sr. The molten state tests failed due to phase transformation of the zeolite structure when in contact with the molten LiCl salt.

  5. Nozzle for electric dispersion reactor

    DOE Patents [OSTI]

    Sisson, W.G.; Harris, M.T.; Scott, T.C.; Basaran, O.A.

    1996-04-02

    A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 5 figs.

  6. Nozzle for electric dispersion reactor

    DOE Patents [OSTI]

    Sisson, W.G.; Basaran, O.A.; Harris, M.T.

    1998-04-14

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

  7. Nozzle for electric dispersion reactor

    DOE Patents [OSTI]

    Sisson, W.G.; Harris, M.T.; Scott, T.C.; Basaran, O.A.

    1998-06-02

    A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 5 figs.

  8. Nozzle for electric dispersion reactor

    DOE Patents [OSTI]

    Sisson, Warren G. (Oak Ridge, TN); Harris, Michael T. (Knoxville, TN); Scott, Timothy C. (Knoxville, TN); Basaran, Osman A. (Oak Ridge, TN)

    1998-01-01

    A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  9. Nozzle for electric dispersion reactor

    DOE Patents [OSTI]

    Sisson, Warren G. (Oak Ridge, TN); Harris, Michael T. (Knoxville, TN); Scott, Timothy C. (Knoxville, TN); Basaran, Osman A. (Oak Ridge, TN)

    1996-01-01

    A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  10. Nozzle for electric dispersion reactor

    DOE Patents [OSTI]

    Sisson, Warren G. (Oak Ridge, TN); Basaran, Osman A. (Oak Ridge, TN); Harris, Michael T. (Knoxville, TN)

    1998-01-01

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  11. Nozzle for electric dispersion reactor

    DOE Patents [OSTI]

    Sisson, Warren G. (Oak Ridge, TN); Basaran, Osman A. (Oak Ridge, TN); Harris, Michael T. (Knoxville, TN)

    1995-01-01

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode.

  12. Nozzle for electric dispersion reactor

    DOE Patents [OSTI]

    Sisson, W.G.; Basaran, O.A.; Harris, M.T.

    1995-11-07

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

  13. DEVELOPMENT OF METHODOLOGY AND FIELD DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE

    SciTech Connect (OSTI)

    Berry, T.; Milliken, C.; Martinez-Rodriguez, M.; Hathcock, D.; Heitkamp, M.

    2012-06-04

    This project developed methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of the fuel storage medium and determine the oxide thickness on the spent fuel basin materials. The overall objective of this project was to determine the amount of time fuel has spent in a storage basin to determine if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations. This project developed and validated forensic tools that can be used to predict the age and condition of spent nuclear fuels stored in liquid basins based on key physical, chemical and microbiological basin characteristics. Key parameters were identified based on a literature review, the parameters were used to design test cells for corrosion analyses, tools were purchased to analyze the key parameters, and these were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The key parameters identified in the literature review included chloride concentration, conductivity, and total organic carbon level. Focus was also placed on aluminum based cladding because of their application to weapons production. The literature review was helpful in identifying important parameters, but relationships between these parameters and corrosion rates were not available. Bench scale test systems were designed, operated, harvested, and analyzed to determine corrosion relationships between water parameters and water conditions, chemistry and microbiological conditions. The data from the bench scale system indicated that corrosion rates were dependent on total organic carbon levels and chloride concentrations. The highest corrosion rates were observed in test cells amended with sediment, a large microbial inoculum and an organic carbon source. A complete characterization test kit was field tested to characterize the SRS L-Area spent fuel basin. The sampling kit consisted of a TOC analyzer, a YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization was done over a two day period in June 2011, and confirmed that the L Area basin is a well operated facility with low corrosion potential.

  14. Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

    SciTech Connect (OSTI)

    Shwageraus, Eugene; Zhao Xianfeng; Driscoll, Michael J.; Hejzlar, Pavel; Kazimi, Mujid S.; Herring, J. Stephen

    2004-07-15

    A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with {sup 235}U is necessary, and the {sup 235}U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO{sub 2}-UO{sub 2}) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of {sup 233}U from the {sup 232}Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the {sup 233}U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible. However, the large power imbalance between the uranium and thorium regions creates several design challenges, such as higher fission gas release and cladding temperature gradients. A reduction of plutonium generation by a factor of 3 in comparison with all-uranium PWR fuel using the same initial {sup 235}U content was estimated. In contrast to homogeneously mixed U-Th fuel, microheterogeneous fuel has a potential for economic performance comparable to the all-UO{sub 2} fuel provided that the microheterogeneous fuel incremental manufacturing costs are negligibly small.

  15. Recovery of manganese oxides from spent alkaline and zinc–carbon batteries. An application as catalysts for VOCs elimination

    SciTech Connect (OSTI)

    Gallegos, María V.; Falco, Lorena R.; Peluso, Miguel A.; Sambeth, Jorge E.; Thomas, Horacio J.

    2013-06-15

    Highlights: • Manganese oxides were synthesized using spent batteries as raw materials. • Spent alkaline and zinc–carbon size AA batteries were used. • A biohydrometallurgical process was employed to bio-lixiviate batteries. • Manganese oxides were active in the oxidation of VOCs (ethanol and heptane). - Abstract: Manganese, in the form of oxide, was recovered from spent alkaline and zinc–carbon batteries employing a biohydrometallurgy process, using a pilot plant consisting in: an air-lift bioreactor (containing an acid-reducing medium produced by an Acidithiobacillus thiooxidans bacteria immobilized on elemental sulfur); a leaching reactor (were battery powder is mixed with the acid-reducing medium) and a recovery reactor. Two different manganese oxides were recovered from the leachate liquor: one of them by electrolysis (EMO) and the other by a chemical precipitation with KMnO{sub 4} solution (CMO). The non-leached solid residue was also studied (RMO). The solids were compared with a MnO{sub x} synthesized in our laboratory. The characterization by XRD, FTIR and XPS reveal the presence of Mn{sub 2}O{sub 3} in the EMO and the CMO samples, together with some Mn{sup 4+} cations. In the solid not extracted by acidic leaching (RMO) the main phase detected was Mn{sub 3}O{sub 4}. The catalytic performance of the oxides was studied in the complete oxidation of ethanol and heptane. Complete conversion of ethanol occurs at 200 °C, while heptane requires more than 400 °C. The CMO has the highest oxide selectivity to CO{sub 2}. The results show that manganese oxides obtained using spent alkaline and zinc–carbon batteries as raw materials, have an interesting performance as catalysts for elimination of VOCs.

  16. Small modular reactors (SMRs) such as the

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Small modular reactors (SMRs) such as the one illustrated in Figure 1 are being considered by the commercial nuclear power industry as an option for more distributed generation and for replace- ment of older fossil fuel generating facilities. SMRs are more compact than operating pressurized water reactors (PWRs), producing from 50 MWe to 200 MWe as compared to 1000 MWe or higher for their full-sized cousins, and are offered as "expandable" units; that is, their modular design allows

  17. Nuclear Reactor Technologies | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Reactor Technologies Nuclear Reactor Technologies TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Small

  18. Photocatalytic reactor

    DOE Patents [OSTI]

    Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

    1999-01-19

    A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

  19. A NOVEL APPROACH TO SPENT FUEL POOL DECOMMISSIONING

    SciTech Connect (OSTI)

    R. L. Demmer

    2011-04-01

    The Idaho National Laboratory (INL) has been at the forefront of developing methods to reduce the cost and schedule of deactivating spent fuel pools (SFP). Several pools have been deactivated at the INL using an underwater approach with divers. These projects provided a basis for the INL cooperation with the Dresden Nuclear Power Station Unit 1 SFP (Exelon Generation Company) deactivation. It represents the first time that a commercial nuclear power plant (NPP) SFP was decommissioned using this underwater coating process. This approach has advantages in many aspects, particularly in reducing airborne contamination and allowing safer, more cost effective deactivation. The INL pioneered underwater coating process was used to decommission three SFPs with a total combined pool volume of over 900,000 gallons. INL provided engineering support and shared project plans to successfully initiate the Dresden project. This report outlines the steps taken by INL and Exelon to decommission SFPs using the underwater coating process. The rationale used to select the underwater coating process and the advantages and disadvantages are described. Special circumstances are also discussed, such as the use of a remotely-operated underwater vehicle to visually and radiologically map the pool areas that were not readily accessible. A larger project, the INTEC-603 SFP in-situ (grouting) deactivation, is reviewed. Several specific areas where special equipment was employed are discussed and a Lessons Learned evaluation is included.

  20. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    SciTech Connect (OSTI)

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J.; Brey, R.F.; Wright, R.N.; Windes, W.F.

    1999-09-03

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.

  1. U.S. Spent Nuclear Fuel Data as of December 31, 2002

    U.S. Energy Information Administration (EIA) Indexed Site

    Latest spent nuclear data Release Date: October 1, 2004 Spent nuclear fuel data is collected by the Energy Information Administration (EIA) for the Office of Civilian Radioactive...

  2. Summary Report of Commercial reactor Criticality Data for Three Mile Island Unit 1

    SciTech Connect (OSTI)

    Larry B. Wimmer

    2001-08-29

    The objective of the ''Summary Report of Commercial Reactor Criticality Data for Three Mile Island Unit I'' is to present the CRC data for the TMI-1 reactor. Results from the CRC evaluations will support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel. These models and their validation are discussed in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000).

  3. Nuclear reactor characteristics and operational history

    Gasoline and Diesel Fuel Update (EIA)

    2. Ownership Data, Table 3. Characteristics and Operational History Table 1. Nuclear Reactor, State, Type, Net Capacity, Generation, and Capacity Factor PDF XLS Plant/Reactor Name Generator ID State Type 2009 Summer Capacity Net MW(e)1 2010 Annual Generation Net MWh2 Capacity Factor Percent3 Arkansas Nuclear One 1 AR PWR 842 6,607,090 90 Arkansas Nuclear One 2 AR PWR 993 8,415,588 97 Beaver Valley 1 PA PWR 892 7,119,413 91 Beaver Valley 2 PA PWR 885 7,874,151 102 Braidwood Generation Station 1

  4. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  5. Secretary Chu Statement on AP1000 Reactor Design Certification | Department

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    of Energy AP1000 Reactor Design Certification Secretary Chu Statement on AP1000 Reactor Design Certification December 22, 2011 - 3:25pm Addthis Washington, D.C. - U.S. Energy Secretary Steven Chu issued the following statement today in support of the Nuclear Regulatory Commission's (NRC) decision to certify Westinghouse Electric's AP1000 nuclear reactor design, a significant step towards constructing a new generation of U.S. nuclear reactors. In February 2010, the Obama Administration

  6. Background radiation measurements at high power research reactors (Journal

    Office of Scientific and Technical Information (OSTI)

    Article) | SciTech Connect Background radiation measurements at high power research reactors Citation Details In-Document Search This content will become publicly available on October 23, 2016 Title: Background radiation measurements at high power research reactors Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino

  7. Method of cleaning a spent fuel assembly

    SciTech Connect (OSTI)

    Chung, D.K.; Jones, C.E. Jr.

    1989-05-09

    A method is described of cleaning a fuel assembly including surfaces thereof prior to decladding, each assembly surface contaminated with a radioactive alkali metal and comprising a plurality of pressurized metallic fuel pins containing a spent fissible material, the method comprising the sequential steps of: (a) placing the fuel assembly in a sealed chamber; (b) passing a heated, inert gas through the chamber to heat the fuel assembly to a temperature sufficient to cause volatilization of the alkali metal but insufficient to rupture the pressurized metal pins; (c) evacuating the chamber to a pressure of less than 0.5 mm of Hg to further enhance volatilization and removal of the alkali metal and maintaining the chamber at that pressure until the decay heat of the fissile materials causes the temperature of the fuel assembly to increase to a level which would be detrimental to the integrity of the metal pins; (d) cooling the fuel assembly by passing a cool, inert gas through the chamber to reduce the temperature of the fuel assembly to a desired level; (e) repeating the evacuation and cooling steps as required to insure removal of substantially all of the radioactive alkali metal from the assembly surface; and (f) recovering the cleaned fuel assembly from the chamber.

  8. Disposition of ORNL's Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Turner, D. W.; DeMonia, B. C.; Horton, L. L.

    2002-02-26

    This paper describes the process of retrieving, repackaging, and preparing Oak Ridge spent nuclear fuel (SNF) for off-site disposition. The objective of the Oak Ridge SNF Project is to safely, reliably, and efficiently manage SNF that is stored on the Oak Ridge Reservation until it can be shipped off-site. The project required development of several unique processes and the design and fabrication of special equipment to enable the successful retrieval, transfer, and repackaging of Oak Ridge SNF. SNF was retrieved and transferred to a hot cell for repackaging. After retrieval of SNF packages, the storage positions were decontaminated and stainless steel liners were installed to resolve the vulnerability of water infiltration. Each repackaged SNF canister has been transferred from the hot cell back to dry storage until off-site shipments can be made. Three shipments of aluminum-clad SNF were made to the Savannah River Site (SRS), and five shipments of non-aluminum-clad SNF are planned to the Idaho National Engineering and Environmental Laboratory (INEEL). Through the integrated cooperation of several organizations including the U.S. Department of Energy (DOE), Bechtel Jacobs Company LLC (BJC), Oak Ridge National Laboratory (ORNL), and various subcontractors, preparations for the disposition of SNF in Oak Ridge have been performed in a safe and successful manner.

  9. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    SciTech Connect (OSTI)

    Blomeke, J.O.

    2001-07-10

    The industrial facilities that have been built or are under construction in France, the United Kingdom, Sweden, and West Germany to handle light-water reactor (LWR) spent fuel and canisters of vitrified high-level waste before ultimate disposal are described and illustrated with drawings and photographs. Published information on the operating performance of these facilities is also given. This information was assembled for consideration in planning and design of similar equipment and facilities needed for the Federal Waste Management System in the United States.

  10. Analysis of the Reactor Position Independent Monitor (PIM) Diagnostic

    SciTech Connect (OSTI)

    Hayes-Sterbenz, Anna Catherine

    2014-07-17

    In this note I analyze the physics determining the proposed reactor position independent monitor (PIM), which is the ratio (240Pu/239Pu)1/3 × (135Cs/137Cs)1/2. The PIM ratios in any reactor fuel is shown to increase monotonically with the time over which the fuel is irradiated. This is because the Cs ratio determines the neutron flux, while the Pu isotopic ratio is determined by the flux times the irradiation time. If the irradiation time for all fuel rods across the reactor is fixed, the PIM ratio is approximately constant in all rods. However, no information can be extracted from the PIM ratio on Pu isotopics unless both the flux (or Cs ratio) and the irradiation time (from, say, Ru isotopics) are known separately, i.e., the PIM ratio is not a fundamental parameter of any reactor. Thus, unless the PIM ratio has been measured for the specific fuel under interrogation, no information can be deduced from measurements or reactor simulations of PIM ratios in different fuel from the same reactor. However, if a PIM measurement has been in one spent fuel rod from a given reactor, all other rods that are known to have been in the reactor for the same irradiation period can be assumed to have approximately the same PIM ratio.

  11. Decommissioning Plan of the Musashi Reactor and Its Progress

    SciTech Connect (OSTI)

    Tanzawa, Tomio

    2008-01-15

    The Musashi Reactor is a TRIGA-II, tank-type research reactor, as shown in Table 1. The reactor had been operated at maximum thermal power level of 100 kW since first critical, January 30, 1963. Reactor operation was shut down due to small leakage of water from the reactor tank on December 21,1989. After shutdown, investigation of the causes, making plan of repair and discussions on restart or decommissioning had been done. Finally, decision of decommissioning was made in May, 2003. The initial plan of the decommissioning was submitted to the competent authority in January, 2004. Now, the reactor is under decommissioning. The plan of decommissioning and its progress are described. In conclusion: considering the status of undertaking plan of the waste disposal facility for the low level radioactive waste from research reactors, the phased decommissioning was selected for the Musashi Reactor. First phase of the decommissioning activities including the actions of permanent shutdown and delivering the spent nuclear fuels to US DOE was completed.

  12. Small Modular Reactors - SRSCRO

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    River National Laboratory (SRNL) has announced several partnerships to bring refrigerator-sized modular nuclear reactors, known as Small Modular Reactors or SMRs, to the...

  13. ``Sleeping reactor`` irradiations: Shutdown reactor determination of short-lived activation products

    SciTech Connect (OSTI)

    Jerde, E.A.; Glasgow, D.C.

    1998-09-01

    At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux ({phi}) of {approximately} 4 {times} 10{sup 14} n/cm{sup 2} {center_dot} s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of {approximately} 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about {+-} 0.5 s) make irradiations of < 6 s less reliable. Therefore, the determination of these ultra-short-lived species in mixed matrices has not generally been made at HFIR. The authors have found that very short lived activation products can be produced easily during the period after reactor shutdown (SCRAM), but prior to the removal of spent fuel elements. During this 24- to 36-h period (dubbed the ``sleeping reactor``), neutrons are produced in the beryllium reflector by the reaction {sup 9}Be({gamma},n){sup 8}Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to {approximately} 1 {times} 10{sup 10} n/cm{sup 2} {center_dot} s within 1 h. By the time the fuel elements are removed, the flux has dropped to {approximately} 6 {times} 10{sup 8}. Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant.

  14. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  15. Spent fuel test - Climax: technical measurements. Interim report, fiscal year 1981

    SciTech Connect (OSTI)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

    1982-04-30

    The Spent Fuel Test-Climax (SFT-C) is located 420 m below surface in the Climax granite stock on the Nevada Test Site. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized from April to May 1980, initiating the 3- to 5-year-duration test. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Technical objectives of the test led to development of a technical measurements program, which is the subject of this report. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 1-1/2 years of the test on more than 900 channels. Much of the acquired data are now available for analysis and are presented here. Highlights of activities this year include completion of site characterization field work, major modifications to the data acquisition and the management systems, and the addition of instrument evaluation as an explicit objective of the test.

  16. Management Of Hanford KW Basin Knockout Pot Sludge As Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Raymond, R. E. [CH2M HIll Plateau Remediation Company, Richland, WA (United States); Evans, K. M. [AREVA, Avignon (France)

    2012-10-22

    CH2M HILL Plateau Remediation Company (CHPRC) and AREVA Federal Services, LLC (AFS) have been working collaboratively to develop and deploy technologies to remove, transport, and interim store remote-handled sludge from the 10S-K West Reactor Fuel Storage Basin on the U.S. Department of Energy (DOE) Hanford Site near Richland, WA, USA. Two disposal paths exist for the different types of sludge found in the K West (KW) Basin. One path is to be managed as Spent Nuclear Fuel (SNF) with eventual disposal at an SNF at a yet to be licensed repository. The second path will be disposed as remote-handled transuranic (RH-TRU) waste at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, NM. This paper describes the systems developed and executed by the Knockout Pot (KOP) Disposition Subproject for processing and interim storage of the sludge managed as SNF, (i.e., KOP material).

  17. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  18. Spent nuclear fuel project path forward preliminary safety evaluation

    SciTech Connect (OSTI)

    Brehm, J.R.; Crowe, R.D.; Siemer, J.M.; Wojdac, L.F.; Hosler, A.G.

    1995-03-01

    This preliminary safety evaluation (PSE) provides validation of the initial project design criteria for the Spent Nuclear Fuel Project (SNFP) Path Forward for removal of fuel from K Basins.

  19. EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched ...

  20. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    SciTech Connect (OSTI)

    Banerjee, Kaushik; Scaglione, John M

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 keff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  1. TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    Del Cul, G.D.

    2002-10-01

    The initial objective of the work described herein was to identify potential methods and technologies needed to disassemble and dissolve graphite-encapsulated, ceramic-coated gas-cooled-reactor spent fuels so that the oxide fuel components can be separated by means of chemical processing. The purpose of this processing is to recover (1) unburned fuel for recycle, (2) long-lived actinides and fission products for transmutation, and (3) other fission products for disposal in acceptable waste forms. Follow-on objectives were to identify and select the most promising candidate flow sheets for experimental evaluation and demonstration and to address the needs to reduce technical risks of the selected technologies. High-temperature gas-cooled reactors (HTGRs) may be deployed in the next -20 years to (1) enable the use of highly efficient gas turbines for producing electricity and (2) provide high-temperature process heat for use in chemical processes, such as the production of hydrogen for use as clean-burning transportation fuel. Also, HTGR fuels are capable of significantly higher burn-up than light-water-reactor (LWR) fuels or fast-reactor (FR) fuels; thus, the HTGR fuels can be used efficiently for transmutation of fissile materials and long-lived actinides and fission products, thereby reducing the inventory of such hazardous and proliferation-prone materials. The ''deep-burn'' concept, described in this report, is an example of this capability. Processing of spent graphite-encapsulated, ceramic-coated fuels presents challenges different from those of processing spent LWR fuels. LWR fuels are processed commercially in Europe and Japan; however, similar infrastructure is not available for processing of the HTGR fuels. Laboratory studies on the processing of HTGR fuels were performed in the United States in the 1960s and 1970s, but no engineering-scale processes were demonstrated. Currently, new regulations concerning emissions will impact the technologies used in processing the fuel. Potential processing methods will be identified both by a review of the literature regarding the processing of similar fuels and by a reliance on the experience and innovation of the authors. The objective is not to generate an exhaustive list of options but rather to identify a number of potentially practical processing options. These options necessarily take into consideration the chemical characteristics of the entire fuel element and its component parts. Once the practical options are identified, a qualitative assessment of the technical merit and maturity, relative costs, and relative quantity of waste generation will be used to rank the various options. Through this form of analysis, a base-case flow sheet will be identified for further study and development. A fallback flow sheet will also be selected to reduce the overall technical risk of the development plan. To support the base-case flow sheet, a technical development plan will be used to identify the key issues for the highest-rated option(s). In this effort the technical uncertainties will be more fully articulated, and research and development activities will be recommended to reduce the technical risks.

  2. Extraction of uranium from spent fuels using liquefied gases

    SciTech Connect (OSTI)

    Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi

    2007-07-01

    For reprocessing of spent nuclear fuels, a novel method to extract actinides from spent fuel using highly compressed gases, nitrogen dioxide and carbon dioxide was proposed. As a fundamental study, the nitrate conversion with liquefied nitrogen dioxide and the nitrate extraction with supercritical carbon dioxide were demonstrated by using uranium dioxide powder, uranyl nitrate and tri-n-butylphosphate complex in the present study. (authors)

  3. Thermal Cooling Limits of Sbotaged Spent Fuel Pools

    SciTech Connect (OSTI)

    Dr. Thomas G. Hughes; Dr. Thomas F. Lin

    2010-09-10

    To develop the understanding and predictive measures of the post “loss of water inventory” hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others

  4. Safeguards Guidance for Independent Spent Fuel Storage Installations (ISFSI)

    National Nuclear Security Administration (NNSA)

    1) May 2012 Guidance for Independent Spent Fuel Dry Storage Installations Safeguards-by-Design: Guidance for Independent Spent Fuel Dry Storage Installations (ISFSI) Philip Casey Durst (INL Consultant) DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy,

  5. EDI as a Treatment Module in Recycling Spent Rinse Waters

    SciTech Connect (OSTI)

    Donovan, Robert P.; Morrison, Dennis J.

    1999-08-11

    Recycling of the spent rinse water discharged from the wet benches commonly used in semiconductor processing is one tactic for responding to the targets for water usage published in the 1997 National Technology Roadmap for Semiconductors (NTRS). Not only does the NTRS list a target that dramatically reduces total water usage/unit area of silicon manufactured by the industry in the future but for the years 2003 and beyond, the NTRS actually touts goals which would have semiconductor manufacturers drawing less water from a regional water supply per unit area of silicon manufactured than the quantity of ultrapure water (UPW) used in the production of that same silicon. Achieving this latter NTRS target strongly implies more widespread recycling of spent rinse waters at semiconductor manufacturing sites. In spite of the fact that, by most metrics, spent rinse waters are of much higher purity than incoming municipal waters, recycling of these spent rinse waters back into the UPW production plant is not a simple, straightforward task. The rub is that certain of the chemicals used in semiconductor manufacturing, and thus potentially present in trace concentrations (or more) in spent rinse waters, are not found in municipal water supplies and are not necessarily removed by the conventional UPW production sequence used by semiconductor manufacturers. Some of these contaminants, unique to spent rinse waters, may actually foul the resins and membranes of the UPW system, posing a threat to UPW production and potentially even causing a shutdown.

  6. Boraflex panel degradation in spent-fuel storage racks at the South Texas Project

    SciTech Connect (OSTI)

    Hoppes, D.F.

    1996-12-31

    Blackness (neutron absorption) testing was conducted in August 1994 on selected South Texas Project (STP) electric generating station spent-fuel pool (SFP) storage racks as required by the surveillance monitoring program. The tests were performed to determine if gaps had developed in the Boraflex neutron poison material and to determine size and location of any gaps identified. The testing was performed by HOLTEC International using a specially designed logging tool containing a {sup 252}Cf neutron source and four boron trifluoride (BF{sub 3}) thermal neutron detectors.

  7. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  8. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  9. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  10. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    SciTech Connect (OSTI)

    Memmott, M. J.; Stansbury, C.; Taylor, C.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  11. MCNP LWR Core Generator

    SciTech Connect (OSTI)

    Fischer, Noah A.

    2012-08-14

    The reactor core input generator allows for MCNP input files to be tailored to design specifications and generated in seconds. Full reactor models can now easily be created by specifying a small set of parameters and generating an MCNP input for a full reactor core. Axial zoning of the core will allow for density variation in the fuel and moderator, with pin-by-pin fidelity, so that BWR cores can more accurately be modeled. LWR core work in progress: (1) Reflectivity option for specifying 1/4, 1/2, or full core simulation; (2) Axial zoning for moderator densities that vary with height; (3) Generating multiple types of assemblies for different fuel enrichments; and (4) Parameters for specifying BWR box walls. Fuel pin work in progress: (1) Radial and azimuthal zoning for generating further unique materials in fuel rods; (2) Options for specifying different types of fuel for MOX or multiple burn assemblies; (3) Additional options for replacing fuel rods with burnable poison rods; and (4) Control rod/blade modeling.

  12. Update; Sodium advanced fast reactor (SAFR) concept

    SciTech Connect (OSTI)

    Oldenkamp, R.D.; Brunings, J.E. ); Guenther, E. ); Hren, R. )

    1988-01-01

    This paper reports on the sodium advanced fast reactor (SAFR) concept developed by the team of Rockwell International, Combustion Engineering, and Bechtel during the 3-year period extending from January 1985 to December 1987 as one element in the U.S. Department of Energy's Advanced Liquid Metal Reactor Program. In January 1988, the team was expanded to include Duke Engineering and Services, Inc., and the concept development was extended under DOE's Program for Improvement in Advanced Modular LMR Design. The SAFR plant concept employs a 450-MWe pool-type liquid metal cooled reactor as its basic module. The reactor assembly module is a standardized shop-fabricated unit that can be shipped to the plant site by barge for installation. Shop fabrication minimizes nuclear-grade field fabrication and reduces the plant construction schedule. Reactor modules can be used individually or in multiples at a given site to supply the needed generating capacity.

  13. Cooling system for a nuclear reactor

    DOE Patents [OSTI]

    Amtmann, Hans H. (Rancho Santa Fe, CA)

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  14. Advanced burner test reactor preconceptual design report.

    SciTech Connect (OSTI)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

  15. Actinide behavior in the Integral Fast Reactor. Final project report

    SciTech Connect (OSTI)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  16. Generation of energy

    DOE Patents [OSTI]

    Kalina, Alexander I. (12214 Clear Fork, Houston, TX 77077)

    1984-01-01

    A method of generating energy which comprises utilizing relatively lower temperature available heat to effect partial distillation of at least portion of a multicomponent working fluid stream at an intermediate pressure to generate working fluid fractions of differing compositions. The fractions are used to produce at least one main rich solution which is relatively enriched with respect to the lower boiling component, and to produce at least one lean solution which is relatively improverished with respect to the lower boiling component. The pressure of the main rich solution is increased whereafter it is evaporated to produce a charged gaseous main working fluid. The main working fluid is expanded to a low pressure level to release energy. The spent low pressure level working fluid is condensed in a main absorption stage by dissolving with cooling in the lean solution to regenerate an initial working fluid for reuse.

  17. Safety Aspects of Dry Spent Fuel Storage and Spent Fuel Management - 13559

    SciTech Connect (OSTI)

    Botsch, W.; Smalian, S.; Hinterding, P.

    2013-07-01

    Dry storage systems are characterized by passive and inherent safety systems ensuring safety even in case of severe incidents or accidents. After the events of Fukushima, the advantages of such passively and inherently safe dry storage systems have become more and more obvious. As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Following safety aspects must be achieved throughout the storage period: - safe enclosure of radioactive materials, - safe removal of decay heat, - securing nuclear criticality safety, - avoidance of unnecessary radiation exposure. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. Furthermore, transport capability must be guaranteed during and after storage as well as limitation and control of radiation exposure. The safe enclosure of radioactive materials in dry storage casks can be achieved by a double-lid sealing system with surveillance of the sealing system. The safe removal of decay heat must be ensured by the design of the storage containers and the storage facility. The safe confinement of radioactive inventory has to be ensured by mechanical integrity of fuel assembly structures. This is guaranteed, e.g. by maintaining the mechanical integrity of the fuel rods or by additional safety measures for defective fuel rods. In order to ensure nuclear critically safety, possible effects of accidents have also to be taken into consideration. In case of dry storage it might be necessary to exclude the re-positioning of fissile material inside the container and/or neutron moderator exclusion might be taken into account. Unnecessary radiation exposure can be avoided by the cask or canister vault system itself. In Germany dry storage of SF in casks fulfills both transport and storage requirements. Mostly, storage facilities are designed as concrete buildings above the ground, but due to regional constraints, one storage facility has also been built as a rock tunnel. The decay heat is always removed by natural air flow; further technical equipment is not needed. The removal of decay heat and shielding had been modeled and calculated by state-of-the-art computer codes before such a facility has been built. TueV and BAM present their long experience in the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel. Different storage systems and facilities in Germany, Europe and world-wide are compared with respect to the safety aspects mentioned above. Initial points are the safety issues of wet storage of SF, and it is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. The German storage concept for dry storage of SF and HLW is presented and discussed. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

  18. Steam generator support system

    DOE Patents [OSTI]

    Moldenhauer, James E. (Simi Valley, CA)

    1987-01-01

    A support system for connection to an outer surface of a J-shaped steam generator for use with a nuclear reactor or other liquid metal cooled power source. The J-shaped steam generator is mounted with the bent portion at the bottom. An arrangement of elongated rod members provides both horizontal and vertical support for the steam generator. The rod members are interconnected to the steam generator assembly and a support structure in a manner which provides for thermal distortion of the steam generator without the transfer of bending moments to the support structure and in a like manner substantially minimizes forces being transferred between the support structure and the steam generator as a result of seismic disturbances.

  19. Steam generator support system

    DOE Patents [OSTI]

    Moldenhauer, J.E.

    1987-08-25

    A support system for connection to an outer surface of a J-shaped steam generator for use with a nuclear reactor or other liquid metal cooled power source is disclosed. The J-shaped steam generator is mounted with the bent portion at the bottom. An arrangement of elongated rod members provides both horizontal and vertical support for the steam generator. The rod members are interconnected to the steam generator assembly and a support structure in a manner which provides for thermal distortion of the steam generator without the transfer of bending moments to the support structure and in a like manner substantially minimizes forces being transferred between the support structure and the steam generator as a result of seismic disturbances. 4 figs.

  20. US Department of Energy Storage of Spent Fuel and High Level Waste

    SciTech Connect (OSTI)

    Sandra M Birk

    2010-10-01

    ABSTRACT This paper provides an overview of the Department of Energy's (DOE) spent nuclear fuel (SNF) and high level waste (HLW) storage management. Like commercial reactor fuel, DOE's SNF and HLW were destined for the Yucca Mountain repository. In March 2010, the DOE filed a motion with the Nuclear Regulatory Commission (NRC) to withdraw the license application for the repository at Yucca Mountain. A new repository is now decades away. The default for the commercial and DOE research reactor fuel and HLW is on-site storage for the foreseeable future. Though the motion to withdraw the license application and delay opening of a repository signals extended storage, DOE's immediate plans for management of its SNF and HLW remain the same as before Yucca Mountain was designated as the repository, though it has expanded its research and development efforts to ensure safe extended storage. This paper outlines some of the proposed research that DOE is conducting and will use to enhance its storage systems and facilities.

  1. Aerosol reactor production of uniform submicron powders

    DOE Patents [OSTI]

    Flagan, Richard C. (Pasadena, CA); Wu, Jin J. (Pasadena, CA)

    1991-02-19

    A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

  2. The technical and economic impact of minor actinide transmutation in a sodium fast reactor

    SciTech Connect (OSTI)

    Gautier, G. M.; Morin, F.; Dechelette, F.; Sanseigne, E.; Chabert, C.

    2012-07-01

    Within the frame work of the French National Act of June 28, 2006 pertaining to the management of high activity, long-lived radioactive waste, one of the proposed processes consists in transmuting the Minor Actinides (MA) in the radial blankets of a Sodium Fast Reactor (SFR). With this option, we may assess the additional cost of the reactor by comparing two SFR designs, one with no Minor Actinides, and the other involving their transmutation. To perform this exercise, we define a reference design called SFRref, of 1500 MWe that is considered to be representative of the Reactor System. The SFRref mainly features a pool architecture with three pumps, six loops with one steam generator per loop. The reference core is the V2B core that was defined by the CEA a few years ago for the Reactor System. This architecture is designed to meet current safety requirements. In the case of transmutation, for this exercise we consider that the fertile blanket is replaced by two rows of assemblies having either 20% of Minor Actinides or 20% of Americium. The assessment work is performed in two phases. - The first consists in identifying and quantifying the technical differences between the two designs: the reference design without Minor Actinides and the design with Minor Actinides. The main differences are located in the reactor vessel, in the fuel handling system and in the intermediate storage area for spent fuel. An assessment of the availability is also performed so that the impact of the transmutation can be known. - The second consists in making an economic appraisal of the two designs. This work is performed using the CEA's SEMER code. The economic results are shown in relative values. For a transmutation of 20% of MA in the assemblies (S/As) and a hypothesis of 4 kW allowable for the washing device, there is a large external storage demanding a very long cooling time of the S/As. In this case, the economic impact may reach 5% on the capital part of the Levelized Unit Electricity Cost (LUEC). A diminished concentration at 10% of MA, reduces the size of the external storage and the cooling time of the assemblies becomes compatible with the management of the irradiated fuel. Even with a low allowable power for the washing device, the economic impact on the capital cost is less than 2.5%. (authors)

  3. Mission Need Statement: Idaho Spent Fuel Facility Project

    SciTech Connect (OSTI)

    Barbara Beller

    2007-09-01

    Approval is requested based on the information in this Mission Need Statement for The Department of Energy, Idaho Operations Office (DOE-ID) to develop a project in support of the mission established by the Office of Environmental Management to "complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research". DOE-ID requests approval to develop the Idaho Spent Fuel Facility Project that is required to implement the Department of Energy's decision for final disposition of spent nuclear fuel in the Geologic Repository at Yucca Mountain. The capability that is required to prepare Spent Nuclear Fuel for transportation and disposal outside the State of Idaho includes characterization, conditioning, packaging, onsite interim storage, and shipping cask loading to complete shipments by January 1,2035. These capabilities do not currently exist in Idaho.

  4. Information handbook on independent spent fuel storage installations

    SciTech Connect (OSTI)

    Raddatz, M.G.; Waters, M.D.

    1996-12-01

    In this information handbook, the staff of the U.S. Nuclear Regulatory Commission describes (1) background information regarding the licensing and history of independent spent fuel storage installations (ISFSIs), (2) a discussion of the licensing process, (3) a description of all currently approved or certified models of dry cask storage systems (DCSSs), and (4) a description of sites currently storing spent fuel in an ISFSI. Storage of spent fuel at ISFSIs must be in accordance with the provisions of 10 CFR Part 72. The staff has provided this handbook for information purposes only. The accuracy of any information herein is not guaranteed. For verification or for more details, the reader should refer to the respective docket files for each DCSS and ISFSI site. The information in this handbook is current as of September 1, 1996.

  5. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

    SciTech Connect (OSTI)

    Michael A. Pope

    2011-10-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

  6. NUHOMS modular spent-fuel storage system: Performance testing

    SciTech Connect (OSTI)

    Strope, L.A.; McKinnon, M.A. ); Dyksterhouse, D.J.; McLean, J.C. )

    1990-09-01

    This report documents the results of a heat transfer and shielding performance evaluation of the NUTECH HOrizontal MOdular Storage (NUHOMS{reg sign}) System utilized by the Carolina Power and Light Co. (CP L) in an Independent Spent Fuel Storage Installation (ISFSI) licensed by the US Nuclear Regulatory Commission (NRC). The ISFSI is located at CP L's H. B. Robinson Nuclear Plant (HBR) near Hartsville, South Carolina. The demonstration included testing of three modules, first with electric heaters and then with spent fuel. The results indicated that the system was conservatively designed, with all heat transfer and shielding design criteria easily met. 5 refs., 45 figs., 9 tabs.

  7. Method for the regeneration of spent molten zinc chloride

    DOE Patents [OSTI]

    Zielke, Clyde W.; Rosenhoover, William A.

    1981-01-01

    In a process for regenerating spent molten zinc chloride which has been used in the hydrocracking of coal or ash-containing polynuclear aromatic hydrocarbonaceous materials derived therefrom and which contains zinc chloride, zinc oxide, zinc oxide complexes and ash-containing carbonaceous residue, by incinerating the spent molten zinc chloride to vaporize the zinc chloride for subsequent condensation to produce a purified molten zinc chloride: an improvement comprising the use of clay in the incineration zone to suppress the vaporization of metals other than zinc. Optionally water is used in conjunction with the clay to further suppress the vaporization of metals other than zinc.

  8. Spent fuel dissolution studies FY 1991 to 1994

    SciTech Connect (OSTI)

    Gray, W.J.; Wilson, C.N.

    1995-12-01

    Dissolution and transport as a result of groundwater flow are generally accepted as the primary mechanisms by which radionuclides from spent fuel placed in a geologic repository could be released to the biosphere. To help provide a source term for performance assessment calculations, dissolution studies on spent fuel and unirradiated uranium oxides have been conducted over the past few years at Pacific Northwest National Laboratory (PNNL) in support of the Yucca Mountain Site Characterization Project. This report describes work for fiscal years 1991 through 1994. The objectives of these studies and the associated conclusions, which were based on the limited number of tests conducted so far, are described in the following subsections.

  9. Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

    SciTech Connect (OSTI)

    Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.; Baryshnikov, M.V.; Kryukov, O.V.; Khaperskaya, A.V.

    2013-07-01

    REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

  10. Proliferation resistance of small modular reactors fuels

    SciTech Connect (OSTI)

    Polidoro, F.; Parozzi, F.; Fassnacht, F.; Kuett, M.; Englert, M.

    2013-07-01

    In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

  11. Final environmental impact statement. Management of commercially generated radioactive waste. Volume 2. Appendices

    SciTech Connect (OSTI)

    Not Available

    1980-10-01

    This EIS analyzes the significant environmental impacts that could occur if various technologies for management and disposal of high-level and transuranic wastes from commercial nuclear power reactors were to be developed and implemented. This EIS will serve as the environmental input for the decision on which technology, or technologies, will be emphasized in further research and development activities in the commercial waste management program. The action proposed in this EIS is to (1) adopt a national strategy to develop mined geologic repositories for disposal of commercially generated high-level and transuranic radioactive waste (while continuing to examine subseabed and very deep hole disposal as potential backup technologies) and (2) conduct a R and D program to develop such facilities and the necessary technology to ensure the safe long-term containment and isolation of these wastes. The Department has considered in this statement: development of conventionally mined deep geologic repositories for disposal of spent fuel from nuclear power reactors and/or radioactive fuel reprocessing wastes; balanced development of several alternative disposal methods; and no waste disposal action. This volume contains appendices of supplementary data on waste management systems, geologic disposal, radiological standards, radiation dose calculation models, related health effects, baseline ecology, socio-economic conditions, hazard indices, comparison of defense and commercial wastes, design considerations, and wastes from thorium-based fuel cycle alternatives. (DMC)

  12. Petascale algorithms for reactor hydrodynamics.

    SciTech Connect (OSTI)

    Fischer, P.; Lottes, J.; Pointer, W. D.; Siegel, A.

    2008-01-01

    We describe recent algorithmic developments that have enabled large eddy simulations of reactor flows on up to P = 65, 000 processors on the IBM BG/P at the Argonne Leadership Computing Facility. Petascale computing is expected to play a pivotal role in the design and analysis of next-generation nuclear reactors. Argonne's SHARP project is focused on advanced reactor simulation, with a current emphasis on modeling coupled neutronics and thermal-hydraulics (TH). The TH modeling comprises a hierarchy of computational fluid dynamics approaches ranging from detailed turbulence computations, using DNS (direct numerical simulation) and LES (large eddy simulation), to full core analysis based on RANS (Reynolds-averaged Navier-Stokes) and subchannel models. Our initial study is focused on LES of sodium-cooled fast reactor cores. The aim is to leverage petascale platforms at DOE's Leadership Computing Facilities (LCFs) to provide detailed information about heat transfer within the core and to provide baseline data for less expensive RANS and subchannel models.

  13. Attrition reactor system

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  14. Attrition reactor system

    DOE Patents [OSTI]

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  15. B Reactor - Hanford Site

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    War II, B Reactor produced plutonium used in the Trinity Test, as well as for the atomic bomb dropped on Nagasaki, Japan, to end World War II. The reactor was designed and built...

  16. Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

    SciTech Connect (OSTI)

    Trtilek, Radek; Podlaha, Josef [UJV Rez, a. s., Hlavni 130, 25068 Husinec-Rez (Czech Republic)] [UJV Rez, a. s., Hlavni 130, 25068 Husinec-Rez (Czech Republic)

    2013-07-01

    After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the second shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)

  17. Period meter for reactors

    DOE Patents [OSTI]

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  18. Bush Administration Moves Forward to Develop Next Generation...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    ... Japan, South Africa, South Korea, Switzerland, the United Kingdom, and the United States. ... on Sodium Cooled Fast Reactors China and Russia to Join the Generation IV International ...

  19. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  20. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  1. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  2. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  3. Effects of spent fuel types on offsite consequences of hypothetical accidents

    SciTech Connect (OSTI)

    Courtney, J. C.; Dwight, C. C.; Lehto, M. A.

    2000-02-18

    Argonne National Laboratory (ANL) conducts experimental work on the development of waste forms suitable for several types of spent fuel at its facility on the Idaho National Engineering and Environmental Laboratory (INEEL) located 48 km West of Idaho Falls, ID. The objective of this paper is to compare the offsite radiological consequences of hypothetical accidents involving the various types of spent nuclear fuel handled in nonreactor nuclear facilities. The highest offsite total effective dose equivalents (TEDEs) are estimated at a receptor located about 5 km SSE of ANL facilities. Criticality safety considerations limit the amount of enriched uranium and plutonium that could be at risk in any given scenario. Heat generated by decay of fission products and actinides does not limit the masses of spent fuel within any given operation because the minimum time elapsed since fissions occurred in any form is at least five years. At cooling times of this magnitude, fewer than ten radionuclides account for 99% of the projected TEDE at offsite receptors for any credible accident. Elimination of all but the most important nuclides allows rapid assessments of offsite doses with little loss of accuracy. Since the ARF (airborne release fraction), RF (respirable fraction), LPF (leak path fraction) and atmospheric dilution factor ({chi}/Q) can vary by orders of magnitude, it is not productive to consider nuclides that contribute less than a few percent of the total dose. Therefore, only {sup 134}Cs, {sup 137}Cs-{sup 137m}Ba, and the actinides significantly influence the offsite radiological consequences of severe accidents. Even using highly conservative assumptions in estimating radiological consequences, they remain well below current Department of Energy guidelines for highly unlikely accidents.

  4. Separator assembly for use in spent nuclear fuel shipping cask

    DOE Patents [OSTI]

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  5. Pinhole Breaches in Spent Fuel Containers: Some Modeling Considerations

    SciTech Connect (OSTI)

    Casella, Andrew M.; Loyalka, Sudarsham K.; Hanson, Brady D.

    2006-06-04

    This paper replaces PNNL-SA-48024 and incorporates the ANS reviewer's comments, including the change in the title. Numerical methods to solve the equations for gas diffusion through very small breaches in spent fuel containers are presented and compared with previous literature results.

  6. Huizenga leads safety of spent fuel management, radioactive waste

    National Nuclear Security Administration (NNSA)

    management meeting in Vienna | National Nuclear Security Administration Huizenga leads safety of spent fuel management, radioactive waste management meeting in Vienna | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget

  7. President Reagan Calls for a National Spent Fuel Storage Facility |

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration Reagan Calls for a National Spent Fuel Storage Facility | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Countering Nuclear Terrorism About Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Library Bios Congressional Testimony Fact Sheets

  8. Anticipating Potential Waste Acceptance Criteria for Defense Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Rechard, R.P.; Lord, M.E.; Stockman, C.T.; McCurley, R.D.

    1997-12-31

    The Office of Environmental Management of the U.S. Department of Energy is responsible for the safe management and disposal of DOE owned defense spent nuclear fuel and high level waste (DSNF/DHLW). A desirable option, direct disposal of the waste in the potential repository at Yucca Mountain, depends on the final waste acceptance criteria, which will be set by DOE`s Office of Civilian Radioactive Waste Management (OCRWM). However, evolving regulations make it difficult to determine what the final acceptance criteria will be. A method of anticipating waste acceptance criteria is to gain an understanding of the DOE owned waste types and their behavior in a disposal system through a performance assessment and contrast such behavior with characteristics of commercial spent fuel. Preliminary results from such an analysis indicate that releases of 99Tc and 237Np from commercial spent fuel exceed those of the DSNF/DHLW; thus, if commercial spent fuel can meet the waste acceptance criteria, then DSNF can also meet the criteria. In large part, these results are caused by the small percentage of total activity of the DSNF in the repository (1.5%) and regulatory mass (4%), and also because commercial fuel cladding was assumed to provide no protection.

  9. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    SciTech Connect (OSTI)

    Bahri, Che Nor Aniza Che Zainul Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  10. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  11. Downhole hydraulic seismic generator

    DOE Patents [OSTI]

    Gregory, Danny L. (Corrales, NM); Hardee, Harry C. (Albuquerque, NM); Smallwood, David O. (Albuquerque, NM)

    1992-01-01

    A downhole hydraulic seismic generator system for transmitting energy wave vibrations into earth strata surrounding a borehole. The system contains an elongated, unitary housing operably connected to a well head aboveground by support and electrical cabling, and contains clamping apparatus for selectively clamping the housing to the walls of the borehole. The system further comprises a hydraulic oscillator containing a double-actuating piston whose movement is controlled by an electro-servovalve regulating a high pressure hydraulic fluid flow into and out of upper and lower chambers surrounding the piston. The spent hydraulic fluid from the hydraulic oscillator is stored and pumped back into the system to provide high pressure fluid for conducting another run at the same, or a different location within the borehole.

  12. What to Expect When Readying to Move Spent Nuclear Fuel from...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power ...

  13. Assessing the Feasibility of Using Neutron Resonance Transmission Analysis (NRTA) for Assaying Plutonium in Spent Fuel Assemblies

    SciTech Connect (OSTI)

    D. L. Chichester; J. W. Sterbentz

    2012-07-01

    Neutron resonance transmission analysis (NRTA) is an active-interrogation nondestructive assay (NDA) technique capable of assaying spent nuclear fuel to determine plutonium content. Prior experimental work has definitively shown the technique capable of assaying plutonium isotope composition in spent-fuel pins to a precision of approximately 3%, with a spatial resolution of a few millimeters. As a Grand Challenge to investigate NDA options for assaying spent fuel assemblies (SFAs) in the commercial fuel cycle, Idaho National Laboratory has explored the feasibility of using NRTA to assay plutonium in a whole SFA. The goal is to achieve a Pu assay precision of 1%. The NRTA technique uses low-energy neutrons from 0.1-40 eV, at the bottom end of the actinide-resonance range, in a time-of-flight arrangement. Isotopic composition is determined by relating absorption of the incident neutrons to the macroscopic cross-section of the actinides of interest in the material, and then using this information to determine the areal density of the isotopes in the SFA. The neutrons used for NRTA are produced using a pulsed, accelerator-based neutron source. Distinguishable resonances exist for both the plutonium (239,240,241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Based on extensive modeling of the problem using Monte Carlo-based simulation codes, our preliminary results suggest that by rotating an SFA to acquire four symmetric views, sufficient neutron transmission can be achieved to assay a SFA. In this approach multiple scan information for the same pins may also be unfolded to potentially allow the determination of plutonium for sub-regions of the assembly. For a 17 ? 17 pressurized water reactor SFA, a simplistic preliminary analysis indicates the mass of 239Pu may be determined with a precision on the order of 5%, without the need for operator-supplied fuel information or operational histories. This paper will present our work to date on this topic, indicate our preliminary findings for a conceptual assay approach, discuss resilience against spoofing, and outline our future plans for evaluating the NRTA technique for SFA plutonium determination.

  14. Spinning fluids reactor

    DOE Patents [OSTI]

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  15. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  16. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  17. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    SciTech Connect (OSTI)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  18. Conditioning and Repackaging of Spent Radioactive Cs-137 and Co-60 Sealed Sources in Egypt - 13490

    SciTech Connect (OSTI)

    Hasan, M.A.; Selim, Y.T.; El-Zakla, T.

    2013-07-01

    Radioactive Sealed sources (RSSs) are widely use all over the world in medicine, agriculture, industry, research, etc. The accidental misuse and exposure to RSSs has caused significant environmental contamination, serious injuries and many deaths. The high specific activity of the materials in many RSSs means that the spread of as little as microgram quantities can generate significant risk to human health and inhibit the use of buildings and land. Conditioning of such sources is a must to protect humans and environment from the hazard of ionizing radiation and contamination. Conditioning is also increase the security of these sources by decreasing the probability of stolen and/or use in terrorist attacks. According to the law No.7/2010, Egyptian atomic energy authority represented in the hot laboratories and waste management center (centralized waste facility, HLWMC) has the responsibility of collecting, conditioning, storing and management of all types of radioactive waste from all Egyptian territory including spent radioactive sealed sources (SRSSs). This paper explains the conditioning procedures for two of the most common SRSSs, Cs{sup 137} and Co{sup 60} sources which make up more than 90% of the total spent radioactive sealed sources stored in our centralized waste facility as one of the major activities of hot laboratories and waste management center. Conditioning has to meet three main objectives, be acceptable for storage, enable their safe transport, and comply with disposal requirements. (authors)

  19. COBRA-SFS: A thermal-hydraulic analysis code for spent fuel storage and transportation casks

    SciTech Connect (OSTI)

    Michener, T.E.; Rector, D.R.; Cuta, J.M.; Dodge, R.E.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS is a general thermal-hydraulic analysis computer code for prediction of material temperatures and fluid conditions in a wide variety of systems. The code has been validated for analysis of spent fuel storage systems, as part of the Commercial Spent Fuel Management Program of the US Department of Energy. The code solves finite volume equations representing the conservation equations for mass, moment, and energy for an incompressible single-phase heat transfer fluid. The fluid solution is coupled to a finite volume solution of the conduction equation in the solid structure of the system. This document presents a complete description of Cycle 2 of COBRA-SFS, and consists of three main parts. Part 1 describes the conservation equations, constitutive models, and solution methods used in the code. Part 2 presents the User Manual, with guidance on code applications, and complete input instructions. This part also includes a detailed description of the auxiliary code RADGEN, used to generate grey body view factors required as input for radiative heat transfer modeling in the code. Part 3 describes the code structure, platform dependent coding, and program hierarchy. Installation instructions are also given for the various platform versions of the code that are available.

  20. Pebble Bed Reactor review update. Fiscal year 1979 annual report

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    Updated information is presented on the Pebble Bed Reactor (PBR) concept being developed in the Federal Republic of Germany for electricity generation and process heat applications. Information is presented concerning nuclear analysis and core performance, fuel cycle evaluation, reactor internals, and safety and availability.

  1. U.S. Nuclear Generation of Electricity

    U.S. Energy Information Administration (EIA) Indexed Site

    Year Capacity and Generation by State and Reactor 2015 P XLS ... data information in the Annual Energy Review, table 9.2. U. S. Nuclear power plants projected electricity ...

  2. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    SciTech Connect (OSTI)

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  3. Experiments Performed in Substantiation of the Conditioning of BN-350 Spent Cesium Traps Using Lead or Lead-Bismuth Alloy Filling Technology

    SciTech Connect (OSTI)

    O. Romanenko; I. Tazhibaeva; I. Yakovlev; A. Ivanov; D. Wells; A. Herrick; J. Michelbacher; S. Shiganakov

    2009-05-01

    The technology of cleaning cesium radionuclides from sodium coolant at the BN-350 fast reactor was realized in the form of cesium traps of two types: stationary devices connected to the circuit that was to be cleaned and in-core devices installed into the core of reactor when it was not under operation. Carbon-graphite materials were used as sorbents to collect and concentrate radioactive cesium, accumulated in the BN-350 reactor circuits over the decades of their operation, in relatively small volume traps which provided effective radiation-safe conditions for personnel working in proximity to the coolant and equipment of the primary circuit during BN-350 decommissioning. Spent cesium traps, as products unfit for further use, represent solid radioactive wastes. The presence of chemically active sodium, potassium and cesium that are able to react violently with water results in series of problems related to their disposal in the Republic of Kazakhstan. Considering the technology of filling spent cesium traps with lead/lead-bismuth alloy as a priority one for their conditioning, evaluations for safety substantiation were implemented. A set of experiments was implemented aimed at verification of calculations performed in substantiation of the proposed technology: filling a full scale cesium trap mock-up with sodium followed by its draining to determine the optimal regimes of draining; filling bench scale cesium trap mock-ups with sodium and cesium followed by sodium draining and filling with lead or lead-bismuth alloy at different temperatures and filling rates to chose the optimal regimes for filling spent cesium traps; implementation of leachability tests to determine the rate of cesium release from the filling materials into water. This paper provides a description of the experimental program carried out and the main results obtained.

  4. NGNP Reactor Coolant Chemistry Control Study

    SciTech Connect (OSTI)

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  5. Hydrogen and water reactor safety: proceedings

    SciTech Connect (OSTI)

    Not Available

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  6. Exhaust system with emissions storage device and plasma reactor

    DOE Patents [OSTI]

    Hoard, John W. (Livonia, MI)

    1998-01-01

    An exhaust system for a combustion system, comprising a storage device for collecting NO.sub.x, hydrocarbon, or particulate emissions, or mixture of these emissions, and a plasma reactor for destroying the collected emissions is described. After the emission is collected in by the storage device for a period of time, the emission is then destroyed in a non-thermal plasma generated by the plasma reactor. With respect to the direction of flow of the exhaust stream, the storage device must be located before the terminus of the plasma reactor, and it may be located wholly before, overlap with, or be contained within the plasma reactor.

  7. U.S. Spent Nuclear Fuel Data as of December 31,2002 Table 3

    Gasoline and Diesel Fuel Update (EIA)

    permanently discharged in previous years, the historical totals change. BWR Boiling-water reactor; PWR Pressurized-water reactor; HTGR High-temperature gas cooled reactor....

  8. Value recovery from spent alumina-base catalyst

    DOE Patents [OSTI]

    Hyatt, David E. (Northglenn, CO)

    1987-01-01

    A process for the recovery of aluminum and at least one other metal selected from the group consisting of molybdenum, nickel and cobalt from a spent hydrogenation catalyst comprising (1) adding about 1 to 3 parts H.sub.2 SO.sub.4 to each part of spent catalyst in a reaction zone of about 20.degree. to 200.degree. C. under sulfide gas pressure between about 1 and about 35 atmospheres, (2) separating the resultant Al.sub.2 (SO.sub.4).sub.3 solution from the sulfide precipitate in the mixture, (3) oxidizing the remaining sulfide precipitate as an aqueous slurry at about 20.degree. to 200.degree. C. in an oxygen-containing atmosphere at a pressure between about 1 and about 35 atmospheres, (4) separating the slurry to obtain solid molybdic acid and a sulfate liquor containing said at least one metal, and (5) recovering said at least one metal from the sulfate liquor in marketable form.

  9. Removal of arsenic compounds from spent catecholated polymer

    DOE Patents [OSTI]

    Fish, Richard H.

    1985-01-01

    Described is a process for removing arsenic from petroliferous derived liquids by contacting said liquid at an elevated temperature with a divinylbenzene-crosslinked polystyrene having catechol ligands anchored thereon. Also, described is a process for regenerating spent catecholated polystyrene by removal of the arsenic bound to it from contacting petroliferous liquid as described above and involves: a. treating said spent catecholated polystyrene, at a temperature in the range of about 20.degree. to 100.degree. C. with an aqueous solution of at least one carbonate and/or bicarbonate of ammonium, alkali and alkaline earth metals, said solution having a pH between about 8 and 10 and, b. separating the solids and liquids from each other. Preferably the regeneration treatment is in two steps wherein step (a) is carried out with an aqueous alcoholic carbonate solution containing lower alkyl alcohol, and, steps (a) and (b) are repeated using a bicarbonate.

  10. Molten tin reprocessing of spent nuclear fuel elements

    DOE Patents [OSTI]

    Heckman, Richard A.

    1983-01-01

    A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

  11. Levelized Power Generation Cost Codes

    Energy Science and Technology Software Center (OSTI)

    1996-04-30

    LPGC is a set of nine microcomputer programs for estimating power generation costs for large steam-electric power plants. These programs permit rapid evaluation using various sets of economic and technical ground rules. The levelized power generation costs calculated may be used to compare the relative economics of nuclear and coal-fired plants based on life-cycle costs. Cost calculations include capital investment cost, operation and maintenance cost, fuel cycle cost, decommissioning cost, and total levelized power generationmore » cost. These programs can be used for quick analyses of power generation costs using alternative economic parameters, such as interest rate, escalation rate, inflation rate, plant lead times, capacity factor, fuel prices, etc. The two major types of electric generating plants considered are pressurized water reactor (PWR) and pulverized coal-fired plants. Data are also provided for the Large Scale Prototype Breeder (LSPB) type liquid metal reactor.« less

  12. The Production & Generation of Radionuclides from Deuterium-Deuterium...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    invention is for the generation and production of radionuclides generated by D-D and D-T nuclear reactions in fusion reactors and fusion devices for medical, industry and research...

  13. Reactor Controllability of 3-Region-Core Molten Salt Reactor System - A Study on Load Following Capability

    SciTech Connect (OSTI)

    Takahisa Yamamoto; Koshi Mitachi; Masatoshi Nishio

    2006-07-01

    The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of fissile fuels (breeding). Thorium (Th) and uranium-233 ({sup 233}U) are fertile and fissile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650 K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the fission reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. and Shimazu et al. developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to consider the effect of fuel salt flow. (authors)

  14. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  15. Tomorrow's Nuclear Reactors are Closer Than You Think | Department of

    Energy Savers [EERE]

    Energy Tomorrow's Nuclear Reactors are Closer Than You Think Tomorrow's Nuclear Reactors are Closer Than You Think March 1, 2016 - 1:00pm Addthis Dr. Rachel Slaybaugh is among the new generation of scientists seeking to revolutionize nuclear energy. She is an assistant professor of nuclear engineering at the University of California-Berkeley. | Photo courtesy of UC Berkeley. Dr. Rachel Slaybaugh is among the new generation of scientists seeking to revolutionize nuclear energy. She is an

  16. Handling encapsulated spent fuel in a geologic repository environment

    SciTech Connect (OSTI)

    Ballou, L.B.

    1983-02-01

    In support of the Spent Fuel Test-Climate at the U.S. Department of Energy`s Nevada Test Site, a spent-fuel canister handling system has been designed, deployed, and operated successfully during the past five years. This system transports encapsulated commercial spent-fuel assemblies between the packaging facility and the test site ({similar_to}100 km), transfers the canisters 420 m vertically to and from a geologic storage drift, and emplaces or retrieves the canisters from the storage holes in the floor of the drift. The spent-fuel canisters are maintained in a fully shielded configuration at all times during the handling cycle, permitting manned access at any time for response to any abnormal conditions. All normal operations are conducted by remote control, thus assuring as low as reasonably achievable exposures to operators; specifically, we have had no measurable exposure during 30 canister transfer operations. While not intended to be prototypical of repository handling operations, the system embodies a number of concepts, now demonstrated to be safe, reliable, and economical, which may be very useful in evaluating full-scale repository handling alternatives in the future. Among the potentially significant concepts are: Use of an integral shielding plug to minimize radiation streaming at all transfer interfaces. Hydraulically actuated transfer cask jacking and rotation features to reduce excavation headroom requirements. Use of a dedicated small diameter (0.5 m) drilled shaft for transfer between the surface and repository workings. A wire-line hoisting system with positive emergency braking device which travels with the load. Remotely activated grapples - three used in the system - which are insensitive to load orientation. Rail-mounted underground transfer vehicle operated with no personnel underground.

  17. Adsorption of phenol from aqueous systems onto spent oil shale

    SciTech Connect (OSTI)

    Darwish, N.A.; Halhouli, K.A.; Al-Dhoon, N.M. [Jordan Univ. of Science and Technology, Irbid (Jordan)

    1996-03-01

    To evaluate its ability to remove phenol from aqueous solution, Jordanian {open_quotes}spent{close_quotes} oil shale, an abundant natural resource, has been used in an experimental adsorption study. Equilibrium of the system has been determined at three temperatures: 30, 40, and 55{degrees}C. The resulting experimental equilibrium isotherms are well represented by Frendlich, Langmuir, and Redlich-Peterson isotherms. The relevant parameters for these isotherms, as regressed from the experimental equilibrium data, are presented. Effects of solution pH (in the range of 3-11), in addition to effects of three inorganic salts (Kl, KCl, and NaCl), on the equilibrium isotherms were also investigated. The effects of pH in the presence of KI and NaCl were also investigated for a possible interaction between salts and solution pH. The initial concentration of phenol in the aqueous system studied ranges from 10 to 200 ppm. Experimental results show that while an acidic solution has no effect on the adsorption capacity of spent oil shale to phenol, a highly basic solution reduces its adsorbability. No sound effect was observed for the inorganic salts studied on the adsorption of phenol on spent oil shale. The experimental results show that there is no interaction between the pH of solution and the presence of salts. In spite of its ability to remove phenol, spent oil shale showed a very low equilibrium capacity (of an order of magnitude of 1 mg/g). Should the adsorption capacity of the shale be improved (by different treatment processes, such as grafting, surface conditioning), results of this study will find a direct practical implication in serving as {open_quotes}raw{close_quotes} reference data for comparison purposes.

  18. Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management

    Office of Environmental Management (EM)

    NRC's Integrated Strategy for NRC s Integrated Strategy for Spent Fuel Management Earl Easton 1 U.S. Nuclear Regulatory Commission May 25, 2010 Road to Yucca Mountain * 20+ years of preparation for the licensing i review * DOE application received in June 2008 and accepted for review in September 2008 * President Obama pursues alternatives to Yucca Mountain * DOE motion to withdraw in March 2010 2 * DOE motion to withdraw in March 2010 * Blue Ribbon Commission on America's Nuclear Future 2

  19. Method For Processing Spent (Trn,Zr)N Fuel

    DOE Patents [OSTI]

    Miller, William E. (Naperville, IL); Richmann, Michael K. (Woodridge, IL)

    2004-07-27

    A new process for recycling spent nuclear fuels, in particular, mixed nitrides of transuranic elements and zirconium. The process consists of two electrorefiner cells in series configuration. A transuranic element such as plutonium is reduced at the cathode in the first cell, zirconium at the cathode in the second cell, and nitrogen-15 is released and captured for reuse to make transuranic and zirconium nitrides.

  20. Naval Spent Nuclear Fuel disposal Container System Description Document

    SciTech Connect (OSTI)

    N. E. Pettit

    2001-07-13

    The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the natural barrier will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

  1. Management of Spent Dessicant from Vapour Recovery Dryers

    Office of Environmental Management (EM)

    Management of Spent Desiccant from Vapour Recovery Dryers Armando Antoniazzi Tritium Focus Group Meeting April 2014 www.kinectrics.com http://blog.kinectrics.com Life Cycle Management Solutions for the Electricity Industry *Request information from members of the tritium handling community *Specifically, practices associated with handling and disposal of tritiated molecular sieves *Use of molecular sieves to trap tritiated water is a familiar technology in the tritium handling community

  2. Precursors for the Immobilization of Radioactive Cesium and Strontium from Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Ortega, Luis H.; McDeavitt, Sean M.

    2007-07-01

    Next generation processes for the recycling of spent nuclear fuel are being developed by the US Department of Energy; this includes solvent extraction methods developed to isolate Cs and Sr fission products and immobilize them in a stable storage form. In this study, simulated (i.e., non-radioactive) cesium and strontium bearing liquid wastes were created and treated with carbon, silica, and alumina at 700 deg. C. The overall goal of this 2006 Nuclear Energy Research Initiative (NERI) project is to synthesize candidate ceramics for Cs and Sr storage and characterize their behavior. The initial results described here confirm the formation of stable compounds resembling strontianite (SrCO{sub 3}) and cesium-aluminosilicate (CsAlSi{sub 2}O{sub 4}). (authors)

  3. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    SciTech Connect (OSTI)

    Trellue, Holly Renee; Fugate, Michael Lynn; Tobin, Stephen Joesph

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  4. 105-K Basin material design basis feed description for spent nuclear fuel project facilities

    SciTech Connect (OSTI)

    Praga, A.N.

    1998-01-08

    Revisions 0 and 0A of this document provided estimated chemical and radionuclide inventories of spent nuclear fuel and sludge currently stored within the Hanford Site`s 105-K Basins. This Revision (Rev. 1) incorporates the following changes into Revision 0A: (1) updates the tables to reflect: improved cross section data, a decision to use accountability data as the basis for total Pu, a corrected methodology for selection of the heat generation basis fee, and a revised decay date; (2) adds section 3.3.3.1 to expand the description of the approach used to calculate the inventory values and explain why that approach yields conservative results; (3) changes the pre-irradiation braze beryllium value.

  5. Integrated data base for 1993: US spent fuel and radioactive waste inventories, projections, and characteristics. Revision 9

    SciTech Connect (OSTI)

    Klein, J.A.; Storch, S.N.; Ashline, R.C.

    1994-03-01

    The Integrated Data Base (IDB) Program has compiled historic data on inventories and characteristics of both commercial and DOE spent fuel; also, commercial and U.S. government-owned radioactive wastes through December 31, 1992. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest U.S. Department of Energy/Energy Information Administration (DOE/EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste (HLW), transuranic (TRU), waste, low-level waste (LLW), commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) LLW. For most of these categories, current and projected inventories are given through the calendar-year (CY) 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal.

  6. Integrated Data Base report--1993: U.S. spent nuclear fuel and radioactive waste inventories, projections, and characteristics. Revision 10

    SciTech Connect (OSTI)

    Not Available

    1994-12-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and DOE spent nuclear fuel; also, commercial and US government-owned radioactive wastes through December 31, 1993. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 256 refs., 38 figs., 141 tabs.

  7. Integrated Data Base for 1992: US spent fuel and radioactive waste inventories, projections, and characteristics. Revision 8

    SciTech Connect (OSTI)

    Payton, M. L.; Williams, J. T.; Tolbert-Smith, M.; Klein, J. A.

    1992-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1991. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal.

  8. PUSH-PULL POWER REACTOR

    DOE Patents [OSTI]

    Froman, D.K.

    1959-02-24

    Power generating nuclear reactors of the homogeneous liquid fuel type are discussed. The apparatus utilizes two identical reactors interconnected by conduits through heat exchanging apparatus. Each reactor contains a critical geometry region and a vapor region separated from the critical region by a baffle. When the liquid in the first critical region becomes critical, the vapor pressure above the fuel is increased due to the rise in the temperature until it forces the liquid fuel out of the first critical region through the heat exchanger and into the second critical region, which is at a lower temperature and consequently a lower vapor pressure. The above reaction is repeated in the second critical region and the liquid fuel is forced back into the first critical region. In this manner criticality is achieved alternately in each critical region and power is extracted by the heat exchanger from the liquid fuel passing therethrough. The vapor region and the heat exchanger have a non-critical geometry and reactivity control is effected by conventional control rods in the critical regions.

  9. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani (Karhula, FI)

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  10. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  11. H Reactor - Hanford Site

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    About Us Projects & Facilities H Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area 324 Building 325 Building 400 Area/Fast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim Storage Area Canyon Facilities Cold Test Facility D and DR Reactors Effluent Treatment Facility Environmental

  12. Tokamak reactor first wall

    DOE Patents [OSTI]

    Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

    1984-11-20

    This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

  13. Proliferation Resistant Nuclear Reactor Fuel

    SciTech Connect (OSTI)

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and we posit that the exploration, development, and implementation of intrinsic mechanisms such as discussed here are part of a balanced approach aimed at preventing the misuse of nuclear material for nuclear-energy applications.

  14. Performance assessment for the geological disposal of Deep Burn spent fuel using TTBX

    SciTech Connect (OSTI)

    Van den Akker, B.P.; Ahn, J. [Department of Nuclear Engineering, University of California, Berkeley, CA 94720 (United States)

    2013-07-01

    The behavior of Deep Burn Modular High Temperature Reactor Spent Fuel (DBSF) is investigated in the Yucca Mountain geological repository (YMR) with respect to the annual dose (Sv/yr) delivered to the Reasonably Maximally Exposed Individual (RMEI) from the transport of radionuclides released from the graphite waste matrix. Transport calculations are performed with a novel computer code, TTBX which is capable of modeling transport pathways that pass through heterogeneous geological formations. TTBX is a multi-region extension of the existing single region TTB transport code. Overall the peak annual dose received by the RMEI is seen to be four orders of magnitude lower than the regulatory threshold for exposure, even under pessimistic scenarios. A number of factors contribute to the favorable performance of DBSF. A reduction of one order of magnitude in the peak annual dose received by the RMEI is observed for every order of magnitude increase in the waste matrix lifetime, highlighting the importance of the waste matrix durability and suggesting graphite's utility as a potential waste matrix for the disposal of high-level waste. Furthermore, we see that by incorporating a higher fidelity far-field model the peak annual dose calculated to be received by the RMEI is reduced by two orders of magnitude. By accounting for the heterogeneities of the far field we have simultaneously removed unnecessary conservatisms and improved the fidelity of the transport model. (authors)

  15. FATE Unified Modeling Method for Spent Nuclear Fuel and Sludge Processing, Shipping and Storage - 13405

    SciTech Connect (OSTI)

    Plys, Martin; Burelbach, James; Lee, Sung Jin; Apthorpe, Robert

    2013-07-01

    A unified modeling method applicable to the processing, shipping, and storage of spent nuclear fuel and sludge has been incrementally developed, validated, and applied over a period of about 15 years at the US DOE Hanford site. The software, FATE{sup TM}, provides a consistent framework for a wide dynamic range of common DOE and commercial fuel and waste applications. It has been used during the design phase, for safety and licensing calculations, and offers a graded approach to complex modeling problems encountered at DOE facilities and abroad (e.g., Sellafield). FATE has also been used for commercial power plant evaluations including reactor building fire modeling for fire PRA, evaluation of hydrogen release, transport, and flammability for post-Fukushima vulnerability assessment, and drying of commercial oxide fuel. FATE comprises an integrated set of models for fluid flow, aerosol and contamination release, transport, and deposition, thermal response including chemical reactions, and evaluation of fire and explosion hazards. It is one of few software tools that combine both source term and thermal-hydraulic capability. Practical examples are described below, with consideration of appropriate model complexity and validation. (authors)

  16. COMPLETION OF THE FIRST INTEGRATED SPENT NUCLEAR FUEL TRANSSHIPMENT/INTERIM STORAGE FACILITY IN NW RUSSIA

    SciTech Connect (OSTI)

    Dyer, R.S.; Barnes, E.; Snipes, R.L.; Hoeibraaten, S.; Gran, H.C.; Foshaug, E.; Godunov, V.

    2003-02-27

    Northwest and Far East Russia contain large quantities of unsecured spent nuclear fuel (SNF) from decommissioned submarines that potentially threaten the fragile environments of the surrounding Arctic and North Pacific regions. The majority of the SNF from the Russian Navy, including that from decommissioned nuclear submarines, is currently stored in on-shore and floating storage facilities. Some of the SNF is damaged and stored in an unstable condition. Existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing this amount of fuel. Additional interim storage capacity is required. Most of the existing storage facilities being used in Northwest Russia do not meet health and safety, and physical security requirements. The United States and Norway are currently providing assistance to the Russian Federation (RF) in developing systems for managing these wastes. If these wastes are not properly managed, they could release significant concentrations of radioactivity to these sensitive environments and could become serious global environmental and physical security issues. There are currently three closely-linked trilateral cooperative projects: development of a prototype dual-purpose transport and storage cask for SNF, a cask transshipment interim storage facility, and a fuel drying and cask de-watering system. The prototype cask has been fabricated, successfully tested, and certified. Serial production is now underway in Russia. In addition, the U.S. and Russia are working together to improve the management strategy for nuclear submarine reactor compartments after SNF removal.

  17. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices

    SciTech Connect (OSTI)

    Rechard, R.P.

    1993-12-01

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

  18. Nuclear power generation and fuel cycle report 1996

    SciTech Connect (OSTI)

    1996-10-01

    This report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the worldwide nuclear fuel market. Long term projections of U.S. nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed. A discussion on decommissioning of nuclear power plants is included.

  19. Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor

    Energy Savers [EERE]

    Vessel Manufacturing Within a Factory Environment - Volume 1 | Department of Energy 1 Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel Manufacturing Within a Factory Environment - Volume 1 This study focused on the learning process for the factory built components of the Integrated Reactor Vessel of a generic 100MWe SMR using Pressurized Water Reactor Technology. PDF icon Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel

  20. Recovery of strontium activity from a strontium-82/rubidium-82 generator

    DOE Patents [OSTI]

    Taylor, Wayne A. (Los Alamos, NM); Phillips, Dennis R. (Los Alamos, NM); Sosnowski, Kenneth M. (Freehold Township, NJ)

    1999-10-12

    Strontium-82 is recovered from spent strontium-82/rubidium-82 generators to provide a source of strontium-82 for additional strontium-82/rubidium-82 generators. The process involves stripping of the strontium-82 from used strontium-82/rubidium-82 generators followed by purification of the strontium-82 material to remove additional metal contaminants to desired levels.

  1. BDDR, a new CEA technological and operating reactor database

    SciTech Connect (OSTI)

    Soldevilla, M.; Salmons, S.; Espinosa, B.

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  2. Behavior of 241Am in fast reactor systems - a safeguards perspective

    SciTech Connect (OSTI)

    Beddingfield, David H; Lafleur, Adrienne M

    2009-01-01

    Advanced fuel-cycle developments around the world currently under development are exploring the possibility of disposing of {sup 241}Am from spent fuel recycle processes by burning this material in fast reactors. For safeguards practitioners, this approach could potentially complicate both fresh- and spent-fuel safeguards measurements. The increased ({alpha},n) production in oxide fuels from the {sup 241}Am increases the uncertainty in coincidence assay of Pu in MOX assemblies and will require additional information to make use of totals-based neutron assay of these assemblies. We have studied the behavior of {sup 241}Am-bearing MOX fuel in the fast reactor system and the effect on neutron and gamma-ray source-terms for safeguards measurements. In this paper, we will present the results of simulations of the behavior of {sup 241}Am in a fast breeder reactor system. Because of the increased use of MOX fuel in thermal reactors and advances in fuel-cycle designs aimed at americium disposal in fast reactors, we have undertaken a brief study of the behavior of americium in these systems to better understand the safeguards impacts of these new approaches. In this paper we will examine the behavior of {sup 241}Am in a variety of nuclear systems to provide insight into the safeguards implications of proposed Am disposition schemes.

  3. Method to Reduce Neutron Production in Small Clean Fusion Reactors Inventor

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    --- Samuel A. Cohen | Princeton Plasma Physics Lab Method to Reduce Neutron Production in Small Clean Fusion Reactors Inventor --- Samuel A. Cohen This invention describes a method to reduce neutron production of D-3He-fueled, steady state, small FRC fusion reactors using periodic, co-streaming, energetic ion beams generated by RF. Use of this method will lessen damage to and activation of reactor components and, in doing so, can advance the development of fusion reactors for electrical

  4. REFLECTOR FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  5. EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Transfer and Return of Low-Enriched Uranium Fuel Elements from Idaho National Laboratory to the Research Reactor in Vienna, Austria (DOEEIS-0218F-SA-05 and DOEEIS-0203-SA-05) May ...

  6. Slurry reactor design studies

    SciTech Connect (OSTI)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  7. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  8. REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

    SciTech Connect (OSTI)

    W. Lee Poe, Jr

    1998-10-01

    In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are the same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site data was weighted to develop the regional site, and provides the weighted data used in the CSAR analysis. It is divided into Part 1 that defines time-dependent releases from each regional site, Part 2 that defines transport conditions through the groundwater, and Part 3 that defines transport through surface water and populations using the surface waters for drinking.

  9. APPLICATIONS OF CURRENT TECHNOLOGY FOR CONTINUOUS MONITORING OF SPENT FUEL

    SciTech Connect (OSTI)

    Drayer, R.

    2013-06-09

    Advancements in technology have opened many opportunities to improve upon the current infrastructure surrounding the nuclear fuel cycle. Embedded devices, very small sensors, and wireless technology can be applied to Security, Safety, and Nonproliferation of Spent Nuclear Fuel. Security, separate of current video monitoring systems, can be improved by integrating current wireless technology with a variety of sensors including motion detection, altimeter, accelerometer, and a tagging system. By continually monitoring these sensors, thresholds can be set to sense deviations from nominal values. Then alarms or notifications can be activated as needed. Safety can be improved in several ways. First, human exposure to ionizing radiation can be reduced by using a wireless sensor package on each spent fuel cask to monitor radiation, temperature, humidity, etc. Since the sensor data is monitored remotely operator stay-time is decreased and distance from the spent fuel increased, so the overall radiation exposure is reduced as compared to visual inspections. The second improvement is the ability to monitor continuously rather than periodically. If changes occur to the material, alarm thresholds could be set and notifications made to provide advanced notice of negative data trends. These sensor packages could also record data to be used for scientific evaluation and studies to improve transportation and storage safety. Nonproliferation can be improved for spent fuel transportation and storage by designing an integrated tag that uses current infrastructure for reporting and in an event; tracking can be accomplished using the Iridium satellite system. This technology is similar to GPS but with higher signal strength and penetration power, but lower accuracy. A sensor package can integrate all or some of the above depending on the transportation and storage requirements and regulations. A sensor package can be developed using off the shelf technology and applying it to each specific need. There are products on the market for smart meters, industrial lighting control and home automation that can be applied to the Back End Fuel Cycle. With a little integration and innovation a cost effective solution is achievable.

  10. Method for reprocessing and separating spent nuclear fuels. [Patent application

    DOE Patents [OSTI]

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

    1982-01-19

    Spent nuclear fuels, including actinide fuels, volatile and nonvolatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

  11. Method for reprocessing and separating spent nuclear fuels

    DOE Patents [OSTI]

    Krikorian, Oscar H. (Danville, CA); Grens, John Z. (Livermore, CA); Parrish, Sr., William H. (Walnut Creek, CA)

    1983-01-01

    Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

  12. Closure mechanism and method for spent nuclear fuel canisters

    DOE Patents [OSTI]

    Doman, Marvin J. (Monroeville, PA)

    2004-11-23

    A canister is provided for storing, transporting, and/or disposing of spent nuclear fuel. The canister includes a canister shell, a top shield plug disposed within the canister, and a leak-tight closure arrangement. The closure arrangement includes a shear ring which forms a containment boundary of the canister, and which is welded to the canister shell and top shield plug. An outer seal plate, forming an outer seal, is disposed above the shear ring and is welded to the shield plug and the canister.

  13. SUPPLEMENT ANALYSIS PROPOSED SHIPMENT OF COMMERCIAL SPENT NUCLEAR FUEL

    Energy Savers [EERE]

    SUPPLEMENT ANALYSIS PROPOSED SHIPMENT OF COMMERCIAL SPENT NUCLEAR FUEL TO DOE NATIONAL LABORATORIES FOR RESEARCH AND DEVELOPMENT PURPOSES Office of Nuclear Energy U.S. DEPARTMENT OF ENERGY DECEMBER 2015 DOE/EIS-0203-SA-07 DOE/EIS-0250F-S-1-SA-02 Commercial Fuel Shipment SA DOE/EIS-0203-SA-07 December 2015 CONVERSION FACTORS Metric to English English to Metric Multiply by To get Multiply by To get Area Square kilometers 247.1 Acres Square kilometers 0.3861 Square miles Square meters 10.764 Square

  14. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    SciTech Connect (OSTI)

    Rodney C. Ewing

    2004-10-07

    Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

  15. Evaluation of Salt Coolants for Reactor Applications

    SciTech Connect (OSTI)

    Williams, David F

    2008-01-01

    Molten fluorides were initially developed for use in the nuclear industry as the high-temperature fluid fuel for the Molten Salt Reactor (MSR). The U.S. Department of Energy Office of Nuclear Energy is exploring the use of molten salts as primary and secondary coolants in a new generation of solid-fueled, thermal-spectrum, hightemperature reactors. This paper provides a review of relevant properties for use in evaluation and ranking of salt coolants for high-temperature reactors. Nuclear, physical, and chemical properties were reviewed, and metrics for evaluation are recommended. Chemical properties of the salt were examined to identify factors that affect materials compatibility (i.e., corrosion). Some preliminary consideration of economic factors for the candidate salts is also presented.

  16. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    SciTech Connect (OSTI)

    Francis, Matthew W.; Weber, Charles F.; Pigni, Marco T.; Gauld, Ian C.

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied as much as 10% of the measured values, and 109Ag was consistently over-predicted by as much as 170%. In general, there is a larger uncertainty for modeling radioactive fission products when compared to either the actinides or the stable fission products in SNF. The relative C/E ratios ranged from a few percent for 137Cs up to 60% and 100% for 106Ru and 125Sb, respectively. Limited or no radioactive fission products data exist in the current data sets for reactor types other than PWRs and BWRs. More work is needed in obtaining a greater diversity of radioactive fission product data. While performing this survey, issues leading to inconsistencies in nuclear fission yield data were discovered that specifically impacted the fission product noble gases. Emphasis was given to this legacy data, and corrective actions were taken as described in this report. After the fission yield data were corrected, the stable xenon and krypton fission products were predicted to within 5% of their measurements. However, preliminary results not explicitly given in this report indicate that the relative C/E ratio for the radioactive isotope 85Kr varied as much as 10%. Due to the complex migration and the difficulty in measuring noble gases in the fuel, a more thorough investigation is needed to understand how accurately depletion codes can calculate these gas concentrations.

  17. NUCLEAR REACTOR FUEL SYSTEMS

    DOE Patents [OSTI]

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  18. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  19. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  20. CONTROL FOR NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Lichtenberger, H.V.; Cameron, R.A.

    1959-03-31

    S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.