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Sample records for lwr fuel melt

  1. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton; Lori Braase; Rose Montgomery; Chris Stanek; Robert Montgomery; Lance Snead; Larry Ott; Mike Billone

    2013-09-01

    The Department of Energy (DOE) Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) is conducting research and development on enhanced Accident Tolerant Fuels (ATF) for light water reactors (LWRs). This mission emphasizes the development of novel fuel and cladding concepts to replace the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the ATF research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. The initial effort will focus on implementation in operating reactors or reactors with design certifications. To initiate the development of quantitative metrics for ATR, a LWR Enhanced Accident Tolerant Fuels Metrics Development Workshop was held in October 2012 in Germantown, MD. This paper summarizes the outcome of that workshop and the current status of metrics development for LWR ATF.

  2. Micro reactor physics of MOX fueled LWR

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan)

    2001-09-01

    Upon the background that the LWR fuels become complicated in recent years because of the introduction of high burnup fuels, high density Gd fuels, MOX fuels, the author proposes the Micro Reactor Physics. He intends to investigate the behaviors of neutrons and reactions in a pin rod that have not yet been paid attention. Conventionally the resonance absorption has been evaluated by assuming the uniform effective cross sections in a pin rod. However, due to the self-shielding, the neutron spectrum near the surface of the rod is quite different with that of the center of rod. This fact affects the spatial distributions of Pu isotopes produced during burnup. The spatial distribution of temperature in a rod affects the Doppler coefficient. He solved this problem by the multi-band method. In the case where MOX rods are adjacent with U rods, the spectrum of the current from MOX rods to U rods is different with that of U to MOX. That makes the spatial distribution of azimuthal direction together with that of the infinite lattice. He solved this problem by a cell calculation based on the characteristic method. This report introduces several numerical results of his Micro Reactor Physics. One of the important results is the indication that the conventional Doppler coefficient gives 20% higher (not conservative) value. (K. Tsuchihashi)

  3. Micro reactor physics of MOX fueled LWR

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan)

    2001-09-01

    Upon the background that the LWR fuels become complicated in recent years because of the introduction of high burnup fuels, high density Gd fuels, MOX fuels, the author proposes the Micro Reactor Physics. He intends to investigate the behaviors of neutrons and reactions in a pin rod that have not yet been paid attention. Conventionally the resonance absorption has been evaluated by assuming the uniform effective cross sections in a pin rod. However, due to the self-shielding, the neutron spectrum near the surface of the rod is quite different with that of the center of rod. This fact affects the spatial distributions of Pu isotopes produced during burnup. The spatial distribution of temperature in a rod affects the Doppler coefficient. He solved this problem by the multi-band method. In the case where MOX rods are adjacent with U rods, the spectrum of the current from MOX rods to U rods is different with that of U to MOX. That makes the spatial distribution of azimuthal direction together with that of the infinite lattice. He solved this problem by a cell calculation based on the characteristic method. This report introduces several numerical results of his Micro Reactor Physics. One of the important results is the indication that the conventional Doppler coefficient gives 20% higher (not conservative) value. (K. Tsuchihashi)

  4. Feasibility study on the development of advanced LWR fuel technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others

    1997-07-01

    Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.

  5. Basic Design of a LWR Fuel Compatibility Test Facility (PLUTO)

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Hwan; Chun, Se Young; Kim, Bok Deuk; Park, Jong Kuk; Chun, Tae Hyun; Kim, Hyoung Kyu; Oh, Dong Seok

    2009-04-15

    KAERI is performing a project for developing a compatibility test facility and the relevant technology for an LWR fuel assembly. It includes the compatibility test and the long term wear test for dual fuel assemblies, and the pressure drop test, uplift force test, flow-induced vibration test, damping test, and the debris filtering capability test for a single fuel assembly. This compatibility test facility of the fuel assemblies is named PLUTO from Performance Test Facility for Fuel Assembly Hydraulics and Vibrations. The PLUTO will be basically constructed for a PWR fuel assembly, and it will be considered to test for the fuel assemblies of other reactors.

  6. Development of information management system on LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility.

  7. Thermal conductivity of heterogeneous LWR MOX fuels

    Science.gov (United States)

    Staicu, D.; Barker, M.

    2013-11-01

    It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference

  8. Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-15

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U3Si2 as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U3Si2 fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U3Si2. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

  9. Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-15

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions need to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions and therefore evaluate the qualification of U3Si2 as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U3Si2 fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U3Si2. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

  10. Irradiation effects on thermal properties of LWR hydride fuel

    Science.gov (United States)

    Terrani, Kurt; Balooch, Mehdi; Carpenter, David; Kohse, Gordon; Keiser, Dennis; Meyer, Mitchell; Olander, Donald

    2017-04-01

    Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

  11. Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

    Energy Technology Data Exchange (ETDEWEB)

    Katayama, Y.B.

    1979-04-01

    Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25/sup 0/C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of /sup 137/Cs and /sup 239 +240/Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures.

  12. Nondestructive evaluation of LWR spent fuel shipping casks

    Energy Technology Data Exchange (ETDEWEB)

    Ballard, D.W.

    1978-02-01

    An analysis of nondestructve testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined. A comprehensive bibliography of current NDT methods for cask evaluation in the USA, Great Britain, Japan and West Germany was compiled for this study.

  13. LIFE vs. LWR: End of the Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J C; Blink, J A; Shaw, H F

    2008-10-02

    The worldwide energy consumption in 2003 was 421 quadrillion Btu (Quads), and included 162 quads for oil, 99 quads for natural gas, 100 quads for coal, 27 quads for nuclear energy, and 33 quads for renewable sources. The projected worldwide energy consumption for 2030 is 722 quads, corresponding to an increase of 71% over the consumption in 2003. The projected consumption for 2030 includes 239 quads for oil, 190 quads for natural gas, 196 quads for coal, 35 quads for nuclear energy, and 62 quads for renewable sources [International Energy Outlook, DOE/EIA-0484, Table D1 (2006) p. 133]. The current fleet of light water reactors (LRWs) provides about 20% of current U.S. electricity, and about 16% of current world electricity. The demand for electricity is expected to grow steeply in this century, as the developing world increases its standard of living. With the increasing price for oil and gasoline within the United States, as well as fear that our CO2 production may be driving intolerable global warming, there is growing pressure to move away from oil, natural gas, and coal towards nuclear energy. Although there is a clear need for nuclear energy, issues facing waste disposal have not been adequately dealt with, either domestically or internationally. Better technological approaches, with better public acceptance, are needed. Nuclear power has been criticized on both safety and waste disposal bases. The safety issues are based on the potential for plant damage and environmental effects due to either nuclear criticality excursions or loss of cooling. Redundant safety systems are used to reduce the probability and consequences of these risks for LWRs. LIFE engines are inherently subcritical, reducing the need for systems to control the fission reactivity. LIFE engines also have a fuel type that tolerates much higher temperatures than LWR fuel, and has two safety systems to remove decay heat in the event of loss of coolant or loss of coolant flow. These features of

  14. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

  15. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase

    2013-01-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and

  16. Extending dry storage of spent LWR fuel for 100 years.

    Energy Technology Data Exchange (ETDEWEB)

    Einziger, R. E.

    1998-12-16

    Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72 [1]. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and

  17. Information on the evolution of severe LWR fuel element damage obtained in the CORA program

    Science.gov (United States)

    Schanz, G.; Hagen, S.; Hofmann, P.; Schumacher, G.; Sepold, L.

    1992-06-01

    In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400°C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4C), which are leading to extensive low-temperature melt formation around 1200°C. Interrelations between those basic phenomena, resulting for example in cladding deformation ("flowering") and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ("quenching") are determining the evolution paths of fuel element destruction, which are to be identified. A further important task is the abstraction from mechanistic and microstructural details in order to get a rough classification of damage regimes (temperature and extent), a practicable analytical treatment of the materials behaviour, and a basis for decisions in accident mitigation and management procedures.

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

    Science.gov (United States)

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

    2012-06-01

    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.

  19. Technical Development on Burn-up Credit for Spent LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  20. Technical development on burn-up credit for spent LWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

  1. LWR fuel recycle program quarterly progress report, April--June 1977

    Energy Technology Data Exchange (ETDEWEB)

    Jarrett, J.H. (comp.)

    1977-08-01

    The LWR Fuel Recycle Program is designed to provide information needed by industry to close the back end of the commercial light water reactor (LWR) fuel cycle. Included in this program are activities in support of specific design studies as well as activities with more general application to fuel recycle technology: economic and environmental studies; spent fuel receipt and storage; head-end processes; off-gas treatment; purex process (solvent extraction); finishing processes; waste management; environmental effects; and general support. 11 figures, 7 tables. (DLC)

  2. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  3. Stress analysis for the candidate of lower end fitting of advanced LWR fuel using FEM

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. S.; Moon, Y. C. [Korea University of Technology and Education, Chonan (Korea, Republic of); Kim, H. K. [Korea Nuclear Fuel Company, Taejon (Korea, Republic of)

    2002-10-01

    The geometric modeling has been conducted for the candidate of advanced LWR fuel using the three-dimensional solid modeler. Then the three-dimensional stress analysis using MSC/NASTRAN has been performed. The evaluation for the mechanical integrity of the candidate has been performed based on the stress distribution obtained from the finite elements analysis.

  4. Mathematical Description of Wafer-1, a Three-Dimensional Code for LWR Fuel Performance Analysis

    DEFF Research Database (Denmark)

    Kjær-Pedersen, Niels

    1975-01-01

    This article describes in detail the mathematical formulation used in the WAFER-1 code, which is presently used for three-dimensional analysis of LWR fuel pin performance. The code aims at a prediction of the local stress-strain history in the cladding, especially with regard to the ridging pheno...

  5. LWR nuclear fuel bundle data for use in fuel bundle handling

    Energy Technology Data Exchange (ETDEWEB)

    Weihermiller, W.B.; Allison, G.S.

    1979-09-01

    Although increasing numbers of spent light water reactor (LWR) fuel bundles are moved into storage, no handling equipment is set up to manipulate all of the various types of fuel bundles. This report summarizes fuel bundle information of interest to the designer of such handling equipment. Dimensional descriptions are included with discussions of assembly procedure and manufacturer provisions for handling equipment. No attempt is made to make a complete compilation of dimensional information; the number of fuel bundle designs and design revisions makes it impractical. Because the fuel bundle designs are so varied, any equipment intended for handling all types of bundles will have to be designed with flexibility in mind. Besides the ability to manipulate fuel bundles in space, handling equipment may be required to locate an external surface or to position a cutting operation to avoid breaking a fuel rod pressure boundary. Even with the most sophisticated and flexible handling equipment, some situations will require use of the manufacturers' as-built descriptions of individual fuel bundles.

  6. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.

  7. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

  8. Chemical thermodynamics of complex systems: fission product behavior in LWR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Kohli, R.

    1981-03-01

    A detailed thermodynamic assessment has been made of the chemical reactions of fission products in LWR fuel rods. Using recent thermodynamic data and the in-reactor oxygen potential and temperature range of LWRs, equilibrium thermodynamic calculations were performed for the most plausible reactions of the fission products. The emphasis in this model is on the chemistry of cesium and rubidium and their reactions with the fuel, other fission products, and the zircaloy cladding. The model predictions are discussed for their implications in fuel-cladding interactions.

  9. Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-08-29

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U3Si2 at LWR conditions. The fission gas behavior of U3Si2 can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U3Si2 for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U3Si2 at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U3Si2 as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.

  10. Comment: collection of assay data on isotopic composition in LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Naito, Yoshitaka; Kurosawa, Masayoshi; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Many assay data of LWR spent fuels have been collected from reactors in the world and some of them are already stored in the database SFCOMPO which was constructed on a personal computer IBM PC/AT. On the other hand, Group constant libraries for burnup calculation code ORIGEN-II were generated from the nuclear data file JENDL3.2. These libraries were evaluated by using the assay data in SFCOMPO. (author)

  11. Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-08-29

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions need to be well-understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U3Si2 at LWR conditions. The fission gas behavior of U3Si2 can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U3Si2 for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U3Si2 at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U3Si2 as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.

  12. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    Energy Technology Data Exchange (ETDEWEB)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-10-16

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle.

  13. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  14. FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING

    Directory of Open Access Journals (Sweden)

    WEON-JU KIM

    2013-08-01

    Full Text Available The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade SiCf/SiC composites are briefly reviewed. A CVI-processed SiCf/SiC composite with a PyC or (PyC-SiCn interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.

  15. Development of Voloxidation Process for Treatment of LWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. J.; Jung, I. H.; Shin, J. M. (and others)

    2007-08-15

    The objective of the project is to develop a process which provides a means to recover fuel from the cladding, and to simplify downstream processes by recovering volatile fission products. This work focuses on the process development in three areas ; the measurement and assessment of the release behavior for the volatile and semi-volatile fission products from the voloxidation process, the assessment of techniques to trap and recover gaseous fission products, and the development of process cycles to optimize fuel cladding separation and fuel particle size. High temperature adsorption method of KAERI was adopted in the co-design of OTS for hot experiment in INL. KAERI supplied 6 sets of filter for hot experiment. Three hot experiment in INL hot cell from the 25th of November for two weeks with attaching 4 KAERI staffs had been carried out. The results were promising. For example, trapping efficiency of Cs was 95% and that of I was 99%, etc.

  16. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

  17. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

  18. Nuclear characteristics of Pu fueled LWR and cross section sensitivities

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ., Suita (Japan). Faculty of Engineering

    1998-03-01

    The present status of Pu utilization to thermal reactors in Japan, nuclear characteristics and topics and cross section sensitivities for analysis of Pu fueled thermal reactors are described. As topics we will discuss the spatial self-shielding effect on the Doppler reactivity effect and the cross section sensitivities with the JENDL-3.1 and 3.2 libraries. (author)

  19. Physical and decay characteristics of commercial LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Roddy, J.W.; Claiborne, H.C.; Ashline, R.C.; Johnson, P.J.; Rhyne, B.T.

    1985-10-01

    Information was collected from the literature and from major manufacturers that will be useful in the design and construction of a mined geologic repository for the disposal of light-water-reactor spent fuel. Pertinent data are included on mechanical design characteristics and materials of construction for fuel assemblies and fuel rods and computed values for heat generation rates, radioactivity, and photon and neutron emission rates as a function of time for four reference cases. Calculations were made with the ORIGEN2 computer code for burnups of 27,500 and 40,000 MWd for a typical boiling-water reactor and 33,000 and 60,000 MWd for a typical pressurized-water reactor. The results are presented in figures depicting the individual contributions per metric ton of initial heavy metal for the activation products, fission products, and actinides and their daughters to the radioactivity and thermal power as a function of time. Tables are also presented that list the contribution of each major nuclide to the radioactivity, thermal power, and photons and neutrons emitted for disposal periods from 1 to 100,000 years.

  20. Physical and decay characteristics of commercial LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Roddy, J.W.; Claiborne, H.C.; Ashline, R.C.; Johnson, P.J.; Rhyne, B.T.

    1986-01-01

    Information was collected from the literature and from major manufacturers that will be useful in the design and construction of a mined geologic repository for the disposal of light-water-reactor spent fuel. Pertinent data are included on mechanical design characteristics and materials of construction for fuel assemblies and fuel rods and computed values for heat generation rates, radioactivity, and photon and neutron emission rates as a function of time for four reference cases. Calculations were made with the ORIGEN2 computer code for burnups of 27,500 and 40,000 MWd for a typical boiling-water reactor and 33,000 and 60,000 MWd for a typical pressurized-water reactor. The results are presented in figures depicting the individual contributions per metric ton of initial heavy metal for the activation products, fission products, and actinides and their daughters to the radioactivity and thermal power as a function of time. Tables are also presented that list the contribution of each major nuclide to the radioactivity, thermal power, and photons and neutrons emitted for disposal emitted for disposal periods from 1 to 100,000 years.

  1. Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.; Merckx, K.R.; Holm, J.S.

    1981-01-01

    This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels.

  2. Thermochemical prediction of chemical form distributions of fission products in LWR mixed oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1998-06-01

    Radial distribution of chemical forms of fission products (FPs) in LWR mixed oxide (MOX) fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The amounts of fission products generated in the fuel were calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. The effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that the radial distribution of chemical forms depends strongly on the LHR and the O/M ratio, and is not proportional to that of burnup. (author)

  3. Development of LWR Fuels with Enhanced Accident Tolerance

    Energy Technology Data Exchange (ETDEWEB)

    Lahoda, Edward J. [Westinghouse Electric Company, LLC, Cranberry Woods, PA (United States); Boylan, Frank A. [Westinghouse Electric Company, LLC, Cranberry Woods, PA (United States)

    2015-10-30

    Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U15N and U3Si2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U3Si2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U3Si2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U3Si2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO2 pellets. Pellets and powders of UO2, UN, and U3Si2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO2. Cold spray application of either the amorphous steel or the Ti2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing

  4. Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States); Mei, Zhigang [Argonne National Lab. (ANL), Argonne, IL (United States); Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-10

    Uranium silicide (U3Si2) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U3Si2, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U3Si2 at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuel material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U3Si2 in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U3Si2 at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U3Si2 fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U3Si2 fuel as an accident

  5. Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lippek, H.E.; Schuller, C.R.

    1979-03-01

    A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.

  6. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2B, User's guide to the LWR assemblies data base, Appendix 2C, User's guide to the LWR radiological data base, Appendix 2D, User's guide to the LWR quantities data base

    Energy Technology Data Exchange (ETDEWEB)

    None

    1987-12-01

    This User's Guide for the LWR Assemblies data base system is part of the Characteristics Data Base being developed under the Waste Systems Data Development Program. The objective of the LWR Assemblies data base is to provide access at the personal computer level to information about fuel assemblies used in light-water reactors. The information available is physical descriptions of intact fuel assemblies and radiological descriptions of spent fuel disassembly hardware. The LWR Assemblies data base is a user-oriented menu driven system. Each menu is instructive about its use. Section 5 of this guide provides a sample session with the data base to assist the user.

  7. Storage of LWR (light-water-reactor) spent fuel in air

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, L.E.; Charlot, L.A.; Coleman, J.E. (Pacific Northwest Lab., Richland, WA (USA)); Knoll, R.W. (Johnson Controls, Inc., Madison, WI (USA))

    1989-12-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to determine the oxidation response of light-water-reactor (LWR) spent fuels under conditions appropriate to fuel storage in air. The program is designed to investigate several independent variables that might affect the oxidation behavior of spent fuel. Included are temperature (135 to 230{degree}C), fuel burnup (to about 34 MWd/kgM), reactor type (pressurized and boiling water reactors), moisture level in the air, and the presence of a high gamma field. In continuing tests with declad spent fuel and nonirradiated UO{sub 2} specimens, oxidation rates were monitored by weight-gain measurements and the microstructures of subsamples taken during the weighing intervals were characterized by several analytical methods. The oxidation behavior indicated by weight gain and time to form powder will be reported in Volume III of this series. The characterization results obtained from x-ray diffractometry, transmission electron microscopy, scanning electron microscopy, and Auger electron spectrometry of oxidized fuel samples are presented in this report. 28 refs., 21 figs., 3 tabs.

  8. Analysis of fission gas release in LWR fuel using the BISON code

    Energy Technology Data Exchange (ETDEWEB)

    G. Pastore; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; R.L. Williamson

    2013-09-01

    Recent advances in the development of the finite-element based, multidimensional fuel performance code BISON of Idaho National Laboratory are presented. Specifically, the development, implementation and testing of a new model for the analysis of fission gas behavior in LWR-UO2 fuel during irradiation are summarized. While retaining a physics-based description of the relevant mechanisms, the model is characterized by a level of complexity suitable for application to engineering-scale nuclear fuel analysis and consistent with the uncertainties pertaining to some parameters. The treatment includes the fundamental features of fission gas behavior, among which are gas diffusion and precipitation in fuel grains, growth and coalescence of gas bubbles at grain faces, grain growth and grain boundary sweeping effects, thermal, athermal, and transient gas release. The BISON code incorporating the new model is applied to the simulation of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database, also included in the IAEA coordinated research projects FUMEX-II and FUMEX-III. The comparison of the results with the available experimental data at moderate burn-up is presented, pointing out an encouraging predictive accuracy, without any fitting applied to the model parameters.

  9. Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DelCul, Guillermo Daniel [ORNL; Trowbridge, Lee D [ORNL; Renier, John-Paul [ORNL; Ellis, Ronald James [ORNL; Williams, Kent Alan [ORNL; Spencer, Barry B [ORNL; Collins, Emory D [ORNL

    2009-02-01

    This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.

  10. Uranium nitride as LWR TRISO fuel: Thermodynamic modeling of U-C-N

    Science.gov (United States)

    Besmann, Theodore M.; Shin, Dongwon; Lindemer, Terrence B.

    2012-08-01

    TRISO coated particle fuel is envisioned as a next generation replacement for current urania pellet fuel in LWR applications. To obtain adequate fissile loading the kernel of the TRISO particle will likely need to be UN instead of UO2. In support of the necessary development effort for this new fuel system, an assessment of phase regions of interest in the U-C-N system was undertaken as the fuel will be prepared by the carbothermic reduction of the oxide followed by nitriding, will be in equilibrium with carbon within the TRISO particle, and will react with minor actinides and fission products. The phase equilibria and thermochemistry of the U-C-N system is reviewed, including nitrogen pressure measurements above various phase fields. Measurements were used to confirm an ideal solution model of UN and UC adequately represents the UC1-xNx phase. Agreement with the data was significantly improved by effectively adjusting the Gibbs free energy of UN by +12 kJ/mol. This also required adjustment of the value for the sesquinitride by +17 kJ/mol to obtain agreement with phase equilibria. The resultant model together with reported values for other phases in the system was used to generate isothermal sections of the U-C-N phase diagram. Nitrogen partial pressures were also computed for regions of interest.

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

    Energy Technology Data Exchange (ETDEWEB)

    Fukasawa, T.; Hoshino, K. [Hitachi-GE Nuclear Energy, Ltd, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Takano, M. [Japan Atomic Energy Agency, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Sato, S. [Hokkaido University, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Shimazu, Y. [Fukui University, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan)

    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)

  12. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Kozlowski, Tomasz [Univ. of Illinois, Urbana-Champaign, IL (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States); Trinkle, Dallas [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, Thoms [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); ang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States); Sabharwall, piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-25

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  13. Fission Product Release and Survivability of UN-Kernel LWR TRISO Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, Theodore M [ORNL; Ferber, Mattison K [ORNL; Lin, Hua-Tay [ORNL

    2014-01-01

    A thermomechanical assessment of the LWR application of TRISO fuel with UN kernels was performed. Fission product release under operational and transient temperature conditions was determined by extrapolation from range calculations and limited data from irradiated UN pellets. Both fission recoil and diffusive release were considered and internal particle pressures computed for both 650 and 800 m diameter kernels as a function of buffer layer thickness. These pressures were used in conjunction with a finite element program to compute the radial and tangential stresses generated with a TRISO particle as a function of fluence. Creep and swelling of the inner and outer pyrolytic carbon layers were included in the analyses. A measure of reliability of the TRISO particle was obtained by measuring the probability of survival of the SiC barrier layer and the maximum tensile stress generated in the pyrolytic carbon layers as a function of fluence. These reliability estimates were obtained as functions of the kernel diameter, buffer layer thickness, and pyrolytic carbon layer thickness. The value of the probability of survival at the end of irradiation was inversely proportional to the maximum pressure.

  14. Proposal and analysis of the benchmark problem suite for reactor physics study of LWR next generation fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-10-01

    In order to investigate the calculation accuracy of the nuclear characteristics of LWR next generation fuels, the Research Committee on Reactor Physics organized by JAERI has established the Working Party on Reactor Physics for LWR Next Generation Fuels. The next generation fuels mean the ones aiming for further extended burn-up such as 70 GWd/t over the current design. The Working Party has proposed six benchmark problems, which consists of pin-cell, PWR fuel assembly and BWR fuel assembly geometries loaded with uranium and MOX fuels, respectively. The specifications of the benchmark problem neglect some of the current limitations such as 5 wt% {sup 235}U to achieve the above-mentioned target. Eleven organizations in the Working Party have carried out the analyses of the benchmark problems. As a result, status of accuracy with the current data and method and some problems to be solved in the future were clarified. In this report, details of the benchmark problems, result by each organization, and their comparisons are presented. (author)

  15. Investigation of the melt-down behaviour of massive radial core enclosures during LWR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Hering, W.; Sengpiel, W.; Messainguiral, C. [CEA Centre d' Etudes Nucleaires de Cadarache, 13 - Saint-Paul-les-Durance (France). DRN

    2000-11-01

    At the Institute for Reactor Safety (IRS) of the Forschungszentrum Karlsruhe (FZK) accident analyses were performed for the projected European pressurised water reactor (EPR) up to 1999 using the best estimate severe core damage code SCDAP/RELAP5 (S/R5). From various scenarios investigated with S/R5 the loss-of-offsite power (LOOP) and the 46 cm{sup 2} small break loss of coolant accident (SBLOCA) were selected to be discussed here in some detail. To simulate the heavy reflector (HR) and core barrel (CB) behaviour beyond the capabilities of S/R5 mod 3.2 a detailed stand alone analytical tool (LOWCOR2) was developed and used to determine the time of HR melting, its axial position, the melting velocity and the melt mass. Furthermore, results of MELCOR calculations performed at Siemens/KWU were used for the SBLOCA scenario. The analyses were extended by a feasibility study to find out whether ICARE2 and the commercial FEM code FIDAP are applicable. The axial position of HR and CB melt through strongly depends on the scenario an ranges between 1.0 m and 2.5 m core elevation. The time period to melt down the HR inner edges lasts up to 17 min and a complete melt through of HR and CB is in the order of magnitude of one hour. At melt through time LOWCOR2 calculated a molten steel mass between 10 Mg and 32 Mg and a melt relocation rate of 35 kg/s along the HR inner surface into the core cavity. (orig.) [German] Am Institut fuer Reaktorsicherheit (IRS) des Forschungszentrums Karlsruhe (FZK) wurden bis 1999 Unfallanalysen fuer den projektierten Europaeischen Druckwasser Reaktor (EPR) mit dem ''best estimate'' Kernschmelzcode SCDAP/RELAP5 (S/R5) durchgefuehrt. Von den verschiedenen mit S/R5 untersuchten Szenarien wurden der ''Ausfall der Wechselspannungsnetze'' (LOOP) und das kleine Leck (46 cm{sup 2}) im kalten Strang der Hauptkuehlmittelleitung (SBLOCA) fuer eine ausfuehrlichere Diskussion ausgewaehlt. Um das Verhalten des &apos

  16. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-09-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) or plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  17. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-11-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  18. A Review and Analysis of European Industrial Experience in Handling LWR Spent Fuel and Vitrified High-Level Waste

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Hybrid reprocessing technology of fluoride volatility and solvent extraction. New reprocessing technology, FLUOREX, for LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Fumio [Hitachi Ltd., Ibaraki (Japan)

    2002-11-01

    Hybrid Process of Fluoride Volatility and Solvent Extraction (FLUOREX) has been objected to develop a low cost reprocessing technology for collection of U and MOX (mixture U and Pu) in LWR fuel cycle. Outline, characteristics, technologies, problems and material balance of FLUOREX are explained. LWR spent fuel consists of about 96% U, 1% Pu and about 3% fission products (FP) and minor actinides (MA). FLUOREX method is hybrid system, which isolates about 90% U at high speed and refines by fluoride volatility process and residue about 10% U, Pu, MA and FP are processed by PUREX method after dissolution in acid. The special features are low cost by small type and lightweight, stable without gas Pu and stop of fluorine gas, reducing load of environment, resistance of nuclear proliferation, application of technologies demonstrated and flexible method for fast reactor. Three problems for development are selective fluoridation of U, transportation of oxides in the fluoride residue and dissolution of transported oxides. The preliminary examination of plan showed 800GWD/t processing volume, 200 day/year operation day, about 51 ten-thousand cubic meter volume of plant, about 1/3 Rokkasho reprocessing plant. (S.Y.)

  20. Thermochemical prediction of chemical form distributions of fission products in LWR oxide fuels irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-09-01

    Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. In addition, the effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that approximately 13 mole% of the total amount of Cs compounds exists as CsI and virtually remaining fraction as Cs{sub 2}MoO{sub 4} under the operation condition of LHR below 400 W/cm. On the other hand, when LHR is beyond 400 W/cm under the transient operation condition, its distribution did not change so much from the one under normal operation condition. (author)

  1. Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Burnup LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I. C.; Parks, C. V.

    2000-12-11

    This report has been prepared to review the technical issues important to the prediction of isotopic compositions and source terms for high-burnup, light-water-reactor (LWR) fuel as utilized in the licensing of spent fuel transport and storage systems. The current trend towards higher initial 235U enrichments, more complex assembly designs, and more efficient fuel management schemes has resulted in higher spent fuel burnups than seen in the past. This trend has led to a situation where high-burnup assemblies from operating LWRs now extend beyond the area where available experimental data can be used to validate the computational methods employed to calculate spent fuel inventories and source terms. This report provides a brief review of currently available validation data, including isotopic assays, decay heat measurements, and shielded dose-rate measurements. Potential new sources of experimental data available in the near term are identified. A review of the background issues important to isotopic predictions and some of the perceived technical challenges that high-burnup fuel presents to the current computational methods are discussed. Based on the review, the phenomena that need to be investigated further and the technical issues that require resolution are presented. The methods and data development that may be required to address the possible shortcomings of physics and depletion methods in the high-burnup and high-enrichment regime are also discussed. Finally, a sensitivity analysis methodology is presented. This methodology is currently being investigated at the Oak Ridge National Laboratory as a computational tool to better understand the changing relative significance of the underlying nuclear data in the different enrichment and burnup regimes and to identify the processes that are dominant in the high-burnup regime. The potential application of the sensitivity analysis methodology to help establish a range of applicability for experimental

  2. Development of FUELSIM/MOD0 for the detailed analysis of LWR fuel rod behavior under normal operation conditions with extended burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G.A.; Allison, C.M. [Innovative Systems Software LLC, 1284 South Woodruff, Idaho Falls, ID (United States)

    1999-07-01

    The FUELSIM code is being developed by Innovative Systems Software as part of the international SCDAP Development and Training Program. FUELSIM is being developed as a 'stand-alone' best-estimate fuel behavior code with evaluation modeling options. The long term goal of the code is to predict fuel performance over the full range of conditions from normal operating behavior to severe accident conditions using a combination of models from the FRAPCON-3, FRAP-T6, SCDAP, and MATPRO fuel behavior codes. FUELSIM/MOD0 is the first release of the code and includes models to predict the behavior of LWR fuel rods during normal operating conditions including the influence of extended burnup fuel. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The code models all the important phenomena that occur during normal operating conditions and contains necessary materials properties, water properties, and heat transfer correlations. The code runs on a variety of computers and operating systems including UNIX, LINUX, and Windows NT or 95. (author)

  3. Fuel performance improvement program. Quarterly progress report, April--June 1978. [LWR

    Energy Technology Data Exchange (ETDEWEB)

    Crouthamel, C.E. (comp.)

    1978-07-01

    The Fuel Performance Improvement Program has as its objective the identification and demonstration of fuel concepts with improved power ramp performance. Improved fuels are being sought to allow reduction or elimination of fuel related operating guidelines on nuclear power plants such that the fuel may be power maneuvered within the rates allowed by the system technical specifications. The program contains a combination of out-of-reactor studies, in-reactor experiments and in-reactor demonstrations. Fuel concepts initially being considered include annular pellets, cladding internally coated with graphite and packed-particle fuels. The performance capability of each concept will be compared to a reference fuel of contemporary pellet design by irradiations in the Halden Boiling Water Reactor. Fuel design and process development is being completed and fuel rod fabrication will begin for the Halden test rods and for the first series of in-reactor experiments. The in-reactor demonstrations are being performed in the Big Rock Point reactor to show that the concepts pose no undue risk to commercial operation. Additional concepts may be considered as the result of a state-of-the-technology review of fuel-cladding interaction and assessment of fuel concepts and the out-of-reactor studies. The results of the program will be used to establish the technical bases for design of fuels with improved power ramp performance.

  4. Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

  5. Categorization of failed and damaged spent LWR (light-water reactor) fuel currently in storage

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.

    1987-11-01

    The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs.

  6. Plutonium rock-like fuel LWR nuclear characteristics and transient behavior in accidents

    Energy Technology Data Exchange (ETDEWEB)

    Akie, Hiroshi; Anoda, Yoshinari; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yamaguchi, Chouichi; Sugo, Yukihiro

    1998-03-01

    For the disposition of excess plutonium, rock-like oxide (ROX) fuel systems based on zirconia (ZrO{sub 2}) or thoria (ThO{sub 2}) have been studied. Safety analysis of ROX fueled PWR showed it is necessary to increase Doppler reactivity coefficient and to reduce power peaking factor of zirconia type ROX (Zr-ROX) fueled core. For these improvements, Zr-ROX fuel composition was modified by considering additives of ThO{sub 2}, UO{sub 2} or Er{sub 2}O{sub 3}, and reducing Gd{sub 2}O{sub 3} content. As a result of the modification, comparable, transient behavior to UO{sub 2} fuel PWR was obtained with UO{sub 2}-Er{sub 2}O{sub 3} added Zr-ROX fuel, while the plutonium transmutation capability is slightly reduced. (author)

  7. Assessment of precision gamma scanning for inspecting LWR fuel rods. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, J.R.; Barnes, B.K.; Barnes, M.L.; Hamlin, D.K.; Medina-Ortega, E.G.

    1981-07-01

    Reconstruction of the radial two-dimensional distributions of fission products using projections obtained by nondestructive gamma scanning was evaluated. The filtered backprojection algorithm provided the best reconstruction for simulated gamma-ray sources, as well as for actual irradiated fuel material. Both a low-burnup (11.5 GWd/tU) light-water reactor fuel rod and a high-burnup (179.1 GWd/tU) fast breeder reactor fuel rod were examined using this technique.

  8. Analysis of triso packing fraction and fissile material to DB-MHR using LWR reprocessed fuel

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Clarysson A.M. da; Pereira, Claubia; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Gual, Maritza R., E-mail: clarysson@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: maritzargual@gmail.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2013-07-01

    Gas-cooled and graphite-moderated reactor is being considered the next generation of nuclear power plants because of its characteristic to operate with reprocessed fuel. The typical fuel element consists of a hexagonal block with coolant and fuel channels. The fuel pin is manufactured into compacted ceramic-coated particles (TRISO) which are used to achieve both a high burnup and a high degree of passive safety. This work uses the MCNPX 2.6.0 to simulate the active core of Deep Burn Modular Helium Reactor (DB-MHR) employing PWR (Pressurized Water Reactor) reprocessed fuel. However, before a complete study of DB-MHR fuel cycle and recharge, it is necessary to evaluate the neutronic parameters to some values of TRISO Packing Fractions (PF) and Fissile Material (FM). Each PF and FM combination would generate the best behaviour of neutronic parameters. Therefore, this study configures several PF and FM combinations considering the heterogeneity of TRISO layers and lattice. The results present the best combination of PF and FM values according with the more appropriated behaviour of the neutronic parameters during the burnup. In this way, the optimized combination can be used to future works of MHR fuel cycle and recharge. (author)

  9. Ceria-thoria pellet manufacturing in preparation for plutonia-thoria LWR fuel production

    Science.gov (United States)

    Drera, Saleem S.; Björk, Klara Insulander; Sobieska, Matylda

    2016-10-01

    Thorium dioxide (thoria) has potential to assist in niche roles as fuel for light water reactors (LWRs). One such application for thoria is its use as the fertile component to burn plutonium in a mixed oxide fuel (MOX). Thor Energy and an international consortium are currently irradiating plutonia-thoria (Th-MOX) fuel in an effort to produce data for its licensing basis. During fuel-manufacturing research and development (R&D), surrogate materials were utilized to highlight procedures and build experience. Cerium dioxide (ceria) provides a good surrogate platform to replicate the chemical nature of plutonium dioxide. The project's fuel manufacturing R&D focused on powder metallurgical techniques to ensure manufacturability with the current commercial MOX fuel production infrastructure. The following paper highlights basics of the ceria-thoria fuel production including powder milling, pellet pressing and pellet sintering. Green pellets and sintered pellets were manufactured with average densities of 67.0% and 95.5% that of theoretical density respectively.

  10. Artificial neural network modeling for fission gas release in LWR UO 2 fuel under RIA conditions

    Science.gov (United States)

    Koo, Yang-Hyun; Oh, Jae-Yong; Lee, Byung-Ho; Tahk, Young-Wook; Song, Kun-Woo

    2010-10-01

    A fission gas release (FGR) model was developed by using an artificial neural network method to predict fission gas release in UO 2 fuel under reactivity initiated accident (RIA) conditions. Based on the test data obtained in the CABRI test reactor and nuclear safety research reactor, the model takes into account the effect of the five parameters: pellet average burnup, peak fuel enthalpy, the ratio of peak fuel enthalpy to pulse width, fission gas release during base-irradiation, and grain size of a fuel pellet. The parametric study of the model, producing a physically reasonable trend of FGR for each parameter, shows that the pellet average burnup and the ratio of peak fuel enthalpy to pulse width are two of the most important parameters. Depending on the combination of input values for the five parameters, the application of the model to a fuel rod under typical RIA conditions of light water reactor produces 1.7-14.0% of FGR for the pellet average burnup ranging from 20 to 70 MW d/kg U.

  11. Oxygen potential measurements in high burnup LWR U0 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1995-05-01

    A miniature solid state galvanic cell was used to measure the oxygen potential Δ overlineG( O2) of reactor irradiated U0 2 fuel at different burnups in the range of 28 to ⩾ 150 GWd d/t M. This very high burnup was achieved in the rim region of a fuel with a cross section average burnup of 75 GWd d/t M. The fuels had different enrichments and therefore different contributions of fission of 235U and 239Pu. The temperature range covered was 900 to 1350 K. None of the fuels showed a significant oxidation. Rather, if allowance is made for the dissolved rare earth fission products and the Pu formed during irradiation, some of the fuels were very slightly substoichiometric and the highest possible degree of oxidation corresponded to U0 2.001. In general, the Δ overlineG( O2) at 750°C was about -400 kJ/mol, corresponding to the Δ overlineG( O2) of the reaction Mo + O 2 → MoO 2. The implication of these results which are in contrast to commonly assumed ideas that U0 2 fuel oxidizes due to burnup, are discussed and the importance of the fission product Mo and of the zircaloy clad as oxygen buffers is outlined.

  12. Material Performance of Fully-Ceramic Micro-Encapsulated Fuel under Selected LWR Design Basis Scenarios: Final Report

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; R. S. Sen; M. A. Pope; A. M. Ougouag

    2011-09-01

    The extension to LWRs of the use of Deep-Burn coated particle fuel envisaged for HTRs has been investigated. TRISO coated fuel particles are used in Fully-Ceramic Microencapsulated (FCM) fuel within a SiC matrix rather than the graphite of HTRs. TRISO particles are well characterized for uranium-fueled HTRs. However, operating conditions of LWRs are different from those of HTRs (temperature, neutron energy spectrum, fast fluence levels, power density). Furthermore, the time scales of transient core behavior during accidents are usually much shorter and thus more severe in LWRs. The PASTA code was updated for analysis of stresses in coated particle FCM fuel. The code extensions enable the automatic use of neutronic data (burnup, fast fluence as a function of irradiation time) obtained using the DRAGON neutronics code. An input option for automatic evaluation of temperature rise during anticipated transients was also added. A new thermal model for FCM was incorporated into the code; so-were updated correlations (for pyrocarbon coating layers) suitable to estimating dimensional changes at the high fluence levels attained in LWR DB fuel. Analyses of the FCM fuel using the updated PASTA code under nominal and accident conditions show: (1) Stress levels in SiC-coatings are low for low fission gas release (FGR) fractions of several percent, as based on data of fission gas diffusion in UO{sub 2} kernels. However, the high burnup level of LWR-DB fuel implies that the FGR fraction is more likely to be in the range of 50-100%, similar to Inert Matrix Fuels (IMFs). For this range the predicted stresses and failure fractions of the SiC coating are high for the reference particle design (500 {micro}mm kernel diameter, 100 {micro}mm buffer, 35 {micro}mm IPyC, 35 {micro}mm SiC, 40 {micro}mm OPyC). A conservative case, assuming 100% FGR, 900K fuel temperature and 705 MWd/kg (77% FIMA) fuel burnup, results in a 8.0 x 10{sup -2} failure probability. For a &apos

  13. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...... uses an empirical gas release model combined with a strongly burn-up dependent correction term, developed by the US Nuclear Regulatory Commission. The paper presents the experimental results and the code calculations. It is concluded that the model predictions are in reasonable agreement (within 15...

  14. Three dimensional considerations in thermal-hydraulics of helical cruciform fuel rods for LWR power uprates

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush, E-mail: kshirvan@mit.edu; Kazimi, Mujid S.

    2014-04-01

    Highlights: • We benchmarked the 4 × 4 helical cruciform fuel (HCF) bundle pressure drop experimental data with CFD. • We also benchmarked the 4 × 4 HCF mixing experimental data with CFD. • We derived new friction factors for PWR and BWR designs at PWR and BWR operating conditions from CFD. • We showed the importance of modeling the 3D conduction in HCF in steady state and transient conditions. - Abstract: In order to increase the power density of current and new light water reactor designs, the helical cruciform fuel (HCF) rods have been proposed. The HCF rod is equivalent to a thin cylindrical rod, with 4 fuel containing vanes, wrapped around it. The HCF rods increase the surface area to volume ratio of the fuel and enhance the inter-subchannel mixing due to their helical shape. The rods do not need supporting grids, as they are packed to periodically contact their neighbors along the flow direction, enabling a higher power density in the core. The HCF rods were reported to have the potential to uprate existing PWRs by 45% and BWRs by 20%. In order to quantify the mixing behavior of the HCF rods based on their twist pitch, experiments were previously performed at atmospheric pressures with single phase water in a 4 by 4 HCF and cylindrical rod bundles. In this paper, the experimental results on pressure drop and mixing are benchmarked with computational fluid dynamic (CFD) using steady state the Reynolds average Navier–Stokes (RANS) turbulence model. The sensitivity of the CFD approach to computational domain, mesh size, mesh shape and RANS turbulence models are examined against the experimental conditions. Due to the refined radial velocity profile from the HCF rods twist, the turbulence models showed little sensitivity to the domain. Based on the CFD simulations, the total pressure drops under the PWR and BWR conditions are expected to be about 10% higher than the values previously reported solely from an empirical correlation based on the

  15. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  16. Consequences of postulated losses of LWR spent fuel and plutonium shipping packages at sea

    Energy Technology Data Exchange (ETDEWEB)

    Heaberlin, S.W.; Baker, D.A.; Beyer, C.E.; Friley, J.R.; Mandel, S.; Peterson, P.L.; Sominen, F.A.

    1977-10-01

    The potential consequences of the loss of a large spent fuel cask and of a single 6M plutonium shipping package into the sea for two specific accident cases are estimated. The radiation dose to man through the marine food chain following the loss of undamaged and fire-damaged packages to the continental shelf and in the deep ocean are conservatively estimated. Two failure mechanisms that could lead to release of radioactive material after loss of packages into the ocean have been considered: corrosion and hydrostatic pressure. A third possible mechanism is thermal overpressurization following burial in marine sediments. It was determined that the seals or pressure relief devices on an undamaged spent fuel cask might fail from hydrostatic forces for losses on the continental shelf although some cask designs would retain their integrity at this depth. The population dose to man through the marine food chain following these scenarios has been estimated. The dose estimates are made relating the radioactive material released and the seafood productivity in the region of the release. Doses are based on a one-year consumption of contaminated seafood. The loss of a single plutonium package on the continental shelf is estimated to produce a population dose commitment of less than 250 man-rem for recycle plutonium. The dose commitment to the average individual is less than one millirem. Doses for losses of undamaged casks to the continental shelf and deep ocean and for loss of a fire-damaged cask to the deep ocean were determined to be several orders of magnitude smaller. 22 tables, 10 figures.

  17. Public comments and Task Force responses regarding the environmental survey of the reprocessing and waste management portions of the LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    1977-03-01

    This document contains responses by the NRC Task Force to comments received on the report ''Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle'' (NUREG-0116). These responses are directed at all comments, inclding those received after the close of the comment period. Additional information on the environmental impacts of reprocessing and waste management which has either become available since the publication of NUREG-0116 or which adds requested clarification to the information in that document.

  18. Nuclear Data Uncertainties for Typical LWR Fuel Assemblies and a Simple Reactor Core

    Science.gov (United States)

    Rochman, D.; Leray, O.; Hursin, M.; Ferroukhi, H.; Vasiliev, A.; Aures, A.; Bostelmann, F.; Zwermann, W.; Cabellos, O.; Diez, C. J.; Dyrda, J.; Garcia-Herranz, N.; Castro, E.; van der Marck, S.; Sjöstrand, H.; Hernandez, A.; Fleming, M.; Sublet, J.-Ch.; Fiorito, L.

    2017-01-01

    The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

  19. Storage of LWR spent fuel in air. Volume 3, Results from exposure of spent fuel to fluorine-contaminated air

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, M.E.; Thomas, L.E.

    1995-06-01

    The Behavior of Spent Fuel in Storage (BSFS) Project has conducted research to develop data on spent nuclear fuel (irradiated U0{sub 2}) that could be used to support design, licensing, and operation of dry storage installations. Test Series B conducted by the BSFS Project was designed as a long-term study of the oxidation of spent fuel exposed to air. It was discovered after the exposures were completed in September 1990 that the test specimens had been exposed to an atmosphere of bottled air contaminated with an unknown quantity of fluorine. This exposure resulted in the test specimens reacting with both the oxygen and the fluorine in the oven atmospheres. The apparent source of the fluorine was gamma radiation-induced chemical decomposition of the fluoro-elastomer gaskets used to seal the oven doors. This chemical decomposition apparently released hydrofluoric acid (HF) vapor into the oven atmospheres. Because the Test Series B specimens were exposed to a fluorine-contaminated oven atmosphere and reacted with the fluorine, it is recommended that the Test Series B data not be used to develop time-temperature limits for exposure of spent nuclear fuel to air. This report has been prepared to document Test Series B and present the collected data and observations.

  20. Options for Burning LWR SNF in LIFE Engine

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J

    2008-09-09

    We have pursued two processes in parallel for the burning of LWR SNF in the LIFE engine: (1) solid fuel option and (2) liquid fuel option. Approaches with both are discussed. The assigned Topical Report on liquid fuels is attached.

  1. Environmental survey of the reprocessing and waste management portions of the LWR fuel cycle: a task force report

    Energy Technology Data Exchange (ETDEWEB)

    Bishop, W.P.; Miraglia, F.J. Jr. (eds.)

    1976-10-01

    This Supplement deals with the reprocessing and waste management portions of the nuclear fuel cycle for uranium-fueled reactors. The scope of the report is limited to the illumination of fuel reprocessing and waste management activities, and examination of the environmental impacts caused by these activities on a per-reactor basis. The approach is to select one realistic reprocessing and waste management system and to treat it in enough depth to illuminate the issues involved, the technology available, and the relationships of these to the nuclear fuel cycle in general and its environmental impacts.

  2. Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions

    Science.gov (United States)

    Ott, L. J.; Robb, K. R.; Wang, D.

    2014-05-01

    Following the severe accidents at the Japanese Fukushima Daiichi Nuclear Power Station in 2011, the US Department of Energy initiated research and development on the enhancement of the accident tolerance of light water reactors by the development of fuels/cladding that, in comparison with the standard UO2/Zircaloy (Zr) system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations. Analyses are presented that illustrate the impact of these new candidate fuel/cladding materials on the fuel performance at normal operating conditions and on the reactor system under DB and BDB accident conditions.

  3. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    Science.gov (United States)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  4. Spent fuel data for waste storage programs

    Energy Technology Data Exchange (ETDEWEB)

    Greene, E M

    1980-09-01

    Data on LWR spent fuel were compiled for dissemination to participants in DOE-sponsored waste storage programs. Included are mechanical descriptions of the existing major types of LWR fuel assemblies, spent LWR fuel fission product inventories and decay heat data, and inventories of LWR spent fuel currently in storage, with projections of future quantities.

  5. On the development of LWR fuel analysis code (1). Analysis of the FEMAXI code and proposal of a new model

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-01-01

    This report summarizes the review on the modeling features of FEMAXI code and proposal of a new theoretical equation model of clad creep on the basis of irradiation-induced microstructure change. It was pointed out that plutonium build-up in fuel matrix and non-uniform radial power profile at high burn-up affect significantly fuel behavior through the interconnected effects with such phenomena as clad irradiation-induced creep, fission gas release, fuel thermal conductivity degradation, rim porous band formation and associated fuel swelling. Therefore, these combined effects should be properly incorporated into the models of the FEMAXI code so that the code can carry out numerical analysis at the level of accuracy and elaboration that modern experimental data obtained in test reactors have. Also, the proposed new mechanistic clad creep model has a general formalism which allows the model to be flexibly applied for clad behavior analysis under normal operation conditions and power transients as well for Zr-based clad materials by the use of established out-of-pile mechanical properties. The model has been tested against experimental data, while further verification is needed with specific emphasis on power ramps and transients. (author)

  6. In-pile tests at Karlsruhe of LWR fuel-rod behavior during the heatup phase of a LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.

    1980-01-01

    In order to investigate the influence of a nuclar environment on the mechanisms of fuel-rod failure, in-pile tests simulating the heatup phase of a loss-of-coolant accident in a pressurized-water reactor are being conducted with irradiated and unirradiated short-length single rods in the FR2 reactor at Kernforschungszentrum karlsruhe (Karlsruhe Nuclear Reasearch Center), Federal Republic of Germany, within the Project Nuclear Safety. With nearly 70% of the scheduled tests completed, no such influences have been found. The in-pile burst and deformation data are in good agreement with results from nonnuclear tests with electrically heated fuel-rod simulators. The phenomenon of pellet disintegration, which has been observed in all tests with previously irradiated rods, needs further investigation.

  7. Estimation of the radionuclide inventory in LWR spent fuel assembly structural materials for long-term safety analysis

    Directory of Open Access Journals (Sweden)

    Caruso Stefano

    2016-01-01

    Full Text Available The radionuclide inventory of materials irradiated in a reactor depends on the initial material composition, irradiation history and on the magnitude and spectrum of the neutron flux. The material composition of a fuel assembly structure includes various alloys of Zircaloy, Inconel and stainless steel. The existing impurities in these materials are very important for accurate determination of the activation of all nuclides with a view to assessing the radiological consequences of their geological disposal. In fact, the safety assessments of geological repositories require the average and maximum (in the sense of very conservative inventories of the very long-lived nuclides as input. The purpose of the present work is to describe the methodology applied for determining the activation of these nuclides in fuel assembly structural materials by means of coupled depletion/activation calculations and also to crosscheck the results obtained from two approaches. UO2 and MOX PWR fuels have been simulated using SCALE/TRITON, simultaneously irradiating the fuel region in POWER mode and the cladding region in FLUX mode and aiming to produce binary macro cross-section libraries by applying accurate local neutron spectra in the cladding region as a function of irradiation history that are suitable for activation calculations. The developed activation libraries have been re-employed in a second run using the ORIGEN-S program for a dedicated activation calculation. The axial variation of the neutron flux along the fuel assembly length has also been considered. The SCALE calculations were performed using a 238-group cross-section library, according to the ENDF/B-VII. The results obtained with the ORIGEN-S activation calculations are compared with the results obtained from TRITON via direct irradiation of the cladding, as allowed by the FLUX mode. It is shown that an agreement on the total calculated activities can be found within 55% for MOX and within 22% for

  8. Extension and validation of the TRANSURANUS burn-up model for helium production in high burn-up LWR fuels

    Science.gov (United States)

    Botazzoli, Pietro; Luzzi, Lelio; Brémier, Stephane; Schubert, Arndt; Van Uffelen, Paul; Walker, Clive T.; Haeck, Wim; Goll, Wolfgang

    2011-12-01

    The TRANSURANUS burn-up model (TUBRNP) calculates the local concentration of the actinides, the main fission products, and 4He as a function of the radial position across a fuel rod. In this paper, the improvements in the helium production model as well as the extensions in the simulation of 238-242Pu, 241Am, 243Am and 242-245Cm isotopes are described. Experimental data used for the extended validation include new EPMA measurements of the local concentrations of Nd and Pu and recent SIMS measurements of the radial distributions of Pu, Am and Cm isotopes, both in a 3.5% enriched commercial PWR UO 2 fuel with a burn-up of 80 and 65 MWd/kgHM, respectively. Good agreement has been found between TUBRNP and the experimental data. The analysis has been complemented by detailed neutron transport calculations (VESTA code), and also revealed the need to update the branching ratio for the 241Am(n,γ) 242mAm reaction in typical PWR conditions.

  9. PLUTON: Three-group neutronic code for burnup analysis of isotope generation and depletion in highly irradiated LWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

  10. Effects of heat transfer coefficient treatments on thermal shock fracture prediction for LWR fuel claddings in water quenching

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho; Lee, Jeong Ik; Cheon, Hee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Accurate modeling of thermal shock induced stresses has become ever most important to emerging accident-tolerant ceramic cladding concepts, such as silicon carbide (SiC) and SiC coated zircaloy. Since fractures of ceramic (entirely ceramic or coated) occur by excessive tensile stresses with linear elasticity, modeling transient stress distribution in the material provides a direct indication of the structural integrity. Indeed, even for the current zircaloy cladding material, the oxide layer formed on the surface - where cracks starts to develop upon water quenching - essentially behaves as a brittle ceramic. Hence, enhanced understanding of thermal shock fracture of a brittle material would fundamentally contribute to safety of nuclear reactors for both the current fuel design and that of the coming future. Understanding thermal shock fracture of a brittle material requires heat transfer rate between the solid and the fluid for transient temperature fields of the solid, and structural response of the solid under the obtained transient temperature fields. In water quenching, a solid experiences dynamic time-varying heat transfer rates with phase changes of the fluid over a short quenching period. Yet, such a dynamic change of heat transfer rates during the water quenching transience has been overlooked in assessments of mechanisms, predictability, and uncertainties for thermal shock fracture. Rather, a time-constant heat transfer coefficient, named 'effective heat transfer coefficient' has become a conventional input to thermal shock fracture analysis. No single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic heat transfer coefficient changes with fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials and complete the picture of stress evolution in the quenched solid. The presented result

  11. Assessment of LWR-HTR-GCFR Integrated Cycle

    Directory of Open Access Journals (Sweden)

    Eleonora Bomboni

    2009-01-01

    Full Text Available Preliminary analyses already performed showed that innovative GCRs, both thermal and fast, are very promising candidate to reach the Gen-IV sustainability goal. The integrated LWR-HTR-GCFR basically aims at closing the current nuclear fuel cycle: in principle, thanks to the unique characteristics of Helium coolant reactors, LWR SNF along with DU become valuable material to produce energy. Additionally, burning HMs of LWR SNF means not only a drastic reduction in the Unat demand but also a remarkable decrease in the long-term radiotoxic component of nuclear waste to be geologically stored. This paper focuses on the analyses of the LWR-HTR-GCFR cycle performed by the University of Pisa in the frame of the EU PUMA project (6th FP. Starting from a brief outline of the main characteristics of HTR and GCFR concepts and of the advantages of linking LWR, HTR and GCFR in a symbiotic way, this paper shows the integrated cycle involving a typical LWR (1000 MWe, a PBMR (400 MWth and a GCFR-“E” (2400 MWth. Additionally, a brief overview of the main technological constraints concerning (Pu+MA-based advanced fuels is given, in order to explain and justify the choices made in the framework of the considered cycle. Thereafter, calculations performed and results obtained are described.

  12. Application of Ceramic Bond Coating for Reusable Melting Crucible of Metallic Fuel Slugs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki-Hwan; Song, Hoon; Ko, Young-Mo; Park, Jeong-Yong; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Hong, Ki-Won [Chungnam National University, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel slugs of the driver fuel assembly have been fabricated by injection casting of the fuel alloys under a vacuum state or an inert atmosphere. Traditionally, metal fuel such as a U-Zr alloy system for SFR has been melted in slurry-coated graphite crucibles and cast in slurry-coated quartz tube molds to prevent melt/material interactions. Reactive coatings and porous coatings can be a source of melt contaminations, and fuel losses, respectively. Ceramic Y{sub 2}O{sub 3}, TiC, and TaC coating materials showed no penetration in the protective layer after a melt dipping test. However, the ceramic coating materials showed separations in the coating interface between the substrate and coating layer, or between the coating layer and fuel melt after the dipping test. All plasma-spray coated methods maintained a sound coating state after a dipping test with U-10wt.%Zr melt. A single coating Y{sub 2}O{sub 3}(150) layer and double coating layer of TaC(50)-Y{sub 2}O{sub 3}(100), showed a sound state or little penetration in the protective layer after a dipping test with U-10wt.%Zr-5wt.%RE melt. Injection casting experiments of U-10wt.%Zr and U-10wt.%Zr-5wt.%RE fuel slugs have been performed to investigate the feasibility of a reusable crucible of the metal fuel slugs. U–10wt.%Zr and U–10wt.%Zr–5wt.%RE fuel slugs have been soundly fabricated without significant interactions of the graphite crucibles. Thus, the ceramic plasma-spray coatings are thought to be promising candidate coating methods for a reusable graphite crucible to fabricate metal fuel slugs.

  13. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2E, Physical descriptions of LWR nonfuel assembly hardware, Appendix 2F, User's guide to the LWR nonfuel assembly data base

    Energy Technology Data Exchange (ETDEWEB)

    None

    1987-12-01

    This appendix includes a two to three page Physical Description report for each Non-fuel Assembly (NFA) Hardware item identified from the current data. Information was obtained via subcontracts with these NFA hardware vendors: Babcock and Wildox, Combustion Engineering and Westinghouse. Data for some NFA hardware are not available. For such hardware, the information shown in this report was obtained from the open literature. Efforts to obtain additional information are continuing. NFA hardware can be grouped into six categories: BWR Channels, Control Elements, Guide Tube Plugs/Orifice Rods, Instrumentation, Neutron Poisons, and Neutron Sources. This appendix lists Physical Description reports alphabetically by vendor within each category. Individual Physical Description reports can be generated interactively through the menu-driven LWR Non-Fuel Assembly Hardware Data Base system. These reports can be viewed on the screen, directed to a printer, or saved in a text file for later use. Special reports and compilations of specific data items can be produced on request.

  14. A Model for Molten Fuel-Coolant Interaction during Melt Slumping in a Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sohal, Manohar Singh; Siefken, Larry James

    1999-10-01

    This paper describes a simple fuel melt slumping model to replace the current parametric model in SCDAP/RELAP5. Specifically, a fuel-coolant interaction (FCI) model is developed to analyze the slumping molten fuel, molten fuel breakup, heat transfer to coolant, relocation of the molten droplets, size of a partially solidified particles that settle to the bottom of the lower plenum, and melt-plenum interaction, if any. Considering our objectives, the molten fuel jet breakup model, and fuel droplets Lagrangian model as included in a code TEXAS-V with Eulerian thermal hydraulics for water and steam from SCDAP/RELAP5 were used. The model was assessed with experimental data from MAGICO-2000 tests performed at University of California at Santa Barbara, and FARO Test L-08 performed at Joint Research Center, Ispra, Italy. The comparison was found satisfactory.

  15. Potential for fuel melting and cladding thermal failure during a PCM event in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, M.S.; Croucher, D.W.

    1979-01-01

    The primary concern in nuclear reactor safety is to ensure that no conceivable accident, whether initiated by a failure of the reactor system or by incorrect operation, will lead to a dangerous release of radiation to the environment. A number of hypothesized off-normal power or cooling conditions, generally termed as power-cooling-mismatch (PCM) accidents, are considered in the safety analysis of light water reactors (LWRs). During a PCM accident, film boiling may occur at the cladding surface and cause a rapid temperature increase in the fuel and the cladding, perhaps producing embrittlement of the zircaloy cladding by oxidation. Molten fuel may be produced at the center of the pellets, extrude radially through open cracks in the outer, unmelted portion of the pellet and relocate in the fuel-cladding gap. If the amount of extruded molten fuel is sufficient to establish contact with the cladding, which is at a high temperature during film boiling, the zircaloy cladding may melt. The present work assesses the potential for central fuel melting and thermal failure of the zircaloy cladding due to melting upon being contacted by extruded molten UO/sub 2/-fuel during a PCM event.

  16. Thermal performances and melting risk assessment in a LMFBR fuel pin

    Science.gov (United States)

    Vettraino, F.; Cacciabue, P. C.; Brunelli, F.

    1985-02-01

    A reliable evaluation of fuel temperature is a key safety requirement in the design of the fuel assembly of a nuclear reactor, especially in the case of a LMFBR whose efficient operation requires high thermal performance fuel. The physico-chemical properties such as density, oxygen to metal ratio and thermal conductivity of a typical LMFBR mixed-oxide fuel, which are known to change in a remarkable way under irradiation, strongly affect the temperature profile within the fuel pellet. A statistical analysis of the temperature values in the fuel of the Italian Fast Reactor PEC, has been performed by means of the RSM code (Response Surface Methodology) coupled to a Monte-Carlo Technique (MUP code), in order to demonstrate that the melting risk is substantially negligible.

  17. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 3. Alternatives for interim storage and transportation

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Volume III of the five-volume report contains information on alternatives for interim storage and transportation. Section titles are: interim storage of spent fuel elements; interim storage of chop-leach fuel bundle residues; tank storage of high-level liquid waste; interim storage of solid non-high-level wastes; interim storage of solidified high-level waste; and, transportation alternatives. (JGB)

  18. LWR fuel rod behavior during reactor tests under loss-of-coolant conditions: Results of the FR2 in-pile tests

    Energy Technology Data Exchange (ETDEWEB)

    Karb, E.H.; Sepold, L.; Hofmann, P.; Petersen, C.; Schanz, G.; Zimmermann, H. (Kernforschungszentrum Karlsruhe G.m.b.H. (Germany, F.R.))

    1982-05-01

    Results of the FR2 in-pile tests on fuel rod behavior under loss-of-coolant accident (LOCA) conditions are presented. To investigate the possible influence of a nuclear environment on fuel rod failure mechanisms, unirradiated as well as irradiated (2500 to 35,000 MWd/tsub(U)) PWR-type test fuel rods were exposed to temperature transients simulating the second heatup phase of a LOCA. Loaded by internal overpressure, the cladding ballooned and ruptured. The burst data do not indicate major differences from results obtained out-of-pile with electrically heated fuel rod simulators, and do not show an influence of burnup. The fuel pellets in previously irradiated rods, already cracked during normal operation, crumbled completely in the regions with large cladding deformation. Post-test examinations revealed cladding mechanical behavior and oxidation to be comparable to out-of-pile results, with relatively little fission gas release during the transient.

  19. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 2. Alternatives for waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Volume II of the five-volume report is devoted to the description of alternatives for waste treatment. The discussion is presented under the following section titles: fuel reprocessing modifications; high-level liquid waste solidification; treatment and immobilization of chop-leach fuel bundle residues; treatment of noncombustible solid wastes; treatment of combustible wastes; treatment of non-high-level liquid wastes; recovery of transuranics from non-high-level wastes; immobilization of miscellaneous non-high-level wastes; volatile radioisotope recovery and off-gas treatment; immobilization of volatile radioisotopes; retired facilities (decontamination and decommissioning); and, modification and use of selected fuel reprocessing wastes. (JGB)

  20. Characteristics Data Base: Programmer's guide to the LWR Quantities Data Base

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.E. (DataPhile, Inc., Knoxville, TN (USA)); Moore, R.S. (Automated Sciences Group, Inc., Oak Ridge, TN (USA))

    1990-08-01

    The LWR Quantities Data Base is a menu-driven PC data base developed as part of OCRWM's waste, technical data base on the characteristics of potential repository wastes, which also includes non-LWR spent fuel, high-level and other materials. This programmer's guide completes the documentation for the LWR Quantities Data Base, the user's guide having been published previously. The PC data base itself may be requested from the Oak Ridge National Laboratory, using the order form provided in Volume 1 of publication DOE/RW-0184.

  1. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W [Westinghouse Savannah River Co., Aiken, SC (USA); Hagrman, D L [EG and G Idaho, Inc., Idaho Falls, ID (USA); McClure, P R; Leonard, M T [Science Applications International Corp., Albuquerque, NM (USA)

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  2. Development of melt dilute technology for disposition of aluminum based spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Swift, W.F. [Nuclear Material Management Division Westinghouse Savannah River Company, Savannah River Site Building 707-C, Aiken, SC 29808 (United States)

    2002-07-01

    The US Department of Energy (DOE) has for many years had a program for receipt and disposition of spent nuclear fuels of US origin from research reactors around the world. The research reactor spent nuclear fuel that consists of aluminum alloy composition has historically been returned to the Savannah River Site (SRS) and dispositioned via chemical reprocessing. In 1995, the DOE evaluated a number of alternatives to chemical reprocessing. In 2000, the DOE selected the melt-dilute alternative as the primary disposition path and direct disposal as the backup path. The melt-dilute technology has been developed from lab-scale demonstration up through the construction of a pilot-scale facility. The pilot-scale L-Area Experimental Facility (LEF) has been constructed and is ready for operation. The LEF will be used primarily, to confirm laboratory research on zeolite media for off- gas trapping and remote operability. Favorable results from the LEF are expected to lead to final design of the production melt-dilute facility identified as the Treatment and Storage Facility (TSF). This paper will describe the melt-dilute process and provide a status of the program development. (author)

  3. Preliminary Investigation of Candidate Materials for Use in Accident Resistant Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jason M. Harp; Paul A. Lessing; Blair H. Park; Jakeob Maupin

    2013-09-01

    As part of a Collaborative Research and Development Agreement (CRADA) with industry, Idaho National Laboratory (INL) is investigating several options for accident resistant uranium compounds including silicides, and nitrides for use in future light water reactor (LWR) fuels. This work is part of a larger effort to create accident tolerant fuel forms where changes to the fuel pellets, cladding, and cladding treatment are considered. The goal fuel form should have a resistance to water corrosion comparable to UO2, have an equal to or larger thermal conductivity than uranium dioxide, a melting temperature that allows the material to stay solid under power reactor conditions, and a uranium loading that maintains or improves current LWR power densities. During the course of this research, fuel fabricated at INL will be characterized, irradiated at the INL Advanced Test Reactor, and examined after irradiation at INL facilities to help inform industrial partners on candidate technologies.

  4. ERDA LWR plant technology program: role of government/industry in improving LWR performance

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-07

    Information is presented under the following chapter headings: executive summary; LWR plant outages; LWR plant construction delays and cancellations; programs addressing plant outages, construction delays, and cancellations; need for additional programs to remedy continuing problems; criteria for government role in LWR commercialization; and the proposed government program.

  5. Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

    Science.gov (United States)

    Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

    2017-04-01

    A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

  6. In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures

    Science.gov (United States)

    Miao, Yinbin; Harp, Jason; Mo, Kun; Bhattacharya, Sumit; Baldo, Peter; Yacout, Abdellatif M.

    2017-02-01

    The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

  7. Development of advanced LWR fuel pellet technology

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kun Woo; Kang, K.W.; Kim, K. S.; Yang, J. H.; Kim, Y. M.; Kim, J. H.; Bang, J. B.; Kim, D. H.; Bae, S. O.; Jung, Y. H.; Lee, Y. S.; Kim, B. G.; Kim, S. H

    2000-03-01

    A UO{sub 2} pellet was designed to have a grain size of larger than 12 {mu}m, and a new duplex design that UO{sub 2}-Gd{sub 2}O{sub 3} is in the core and UO{sub 2}-Er{sub 2}O{sub 3} in the periphery was proposed. A master mixing method was developed to make a uniform mixture of UO{sub 2} and additives. The open porosity of UO{sub 2} pellet was reduced by only mixing AUC-UO{sub 2} powder with ADU-UO{sub 2} or milled powder. Duplex compaction tools (die and punch) were designed and fabricated, and duplex compacting procedures were developed to fabricate the duplex BA pellet. In UO{sub 2} sintering, the relations between sintering variables (additive, sintering gas, sintering temperature) and pellet properties (density, grain size, pore size) were experimentally found. The UO{sub 2}-U{sub 3}O{sub 8} powder which is inherently not sinterable to high density could be sintered well with the aid of additives. U{sub 3}O{sub 8} single crystals were added to UO{sub 2} powder, and homogeneous powder mixture was pressed and sintered in a reducing atmosphere. This technology leads to a large-grained pellet of 12-20 {mu}m. In UO{sub 2}-Gd{sub 2}O{sub 3} sintering, the relations between sintering variables (additives, sintering gas) and pellet properties (density, grain size) were experimentally found. The developed technology of fabricating a large-grained UO{sub 2} pellet has been optimized in a lab scale. Pellet properties were investigated in the fields of (1) creep properties, (2) thermal properties, (3) O/M ratios and (4) unit cell lattice. (author)

  8. Evaluation of Codisposal Viability for Melt and Dilute DOE-Owned Fuel

    Energy Technology Data Exchange (ETDEWEB)

    H. Radulescu

    2001-07-31

    There are more than 250 forms of U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment (DOE 2000b, Section 6.6.8). The Melt and Dilute (MD) SNF has been designated as the representative fuel for the high-enriched U-Al fuel group. The MD SNF consists of homogeneous cylindrical ingots with 16.5 in. (419.1 mm) maximum diameter. Two general ingot compositions are considered in the criticality and geochemistry analyses. The first composition consists of 8.2 to 18.2 wt% uranium, enriched at less than 20 wt% U-235 and 0.5 wt% gadolinium, with the balance of the ingot being aluminum. The second composition is identical to the first for uranium and gadolinium, but in this case 2.5 wt% of the ingot is hafnium, with the balance of the ingot being aluminum. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance of fuel from the high-enriched U-Al fuel group for disposal, the important items for the fuel types that are being considered for disposal under the high-enriched U-Al fuel group must be demonstrated to satisfy the conditions determined in this report. The analyses have been performed by following the disposal criticality analysis methodology, which was documented in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000) submitted to the U.S. Nuclear Regulatory Commission. The methodology includes analyzing the geochemical and physical processes that can breach the waste package and degrade the waste forms and other internal components, as well as the structural, thermal, and shielding analyses, and intact and degraded

  9. Development of PIE techniques for irradiated LWR pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-09-01

    For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

  10. Melt-processed anhydrous proton exchange membranes for fuel cell applications

    Energy Technology Data Exchange (ETDEWEB)

    Mokrini, A.; Siu, A.; Diaz, G.; Crites, C.; Robitaille, L. [National Research Council of Canada, Boucherville, PQ (Canada). Industrial Materials Inst.

    2009-07-01

    The current benchmark materials for proton exchange membrane (PEM) fuel cells are perfluorosulfonic acid resins (PFSA) because of their excellent stability and proton conductivity of 0.1 s/cm at 80 degrees C when fully humidified. However their performance decreases significantly at higher temperatures and low humidity. This paper presented the properties of nanocomposite PEMs incorporating a series of anhydrous charge carriers that are viable candidates for making water-free membranes that can operate at temperatures above 120 degrees C. However, the volatility or leaching of these anhydrous charge carriers could prevent them from being successfully used in open electrochemical systems. Therefore, in this study, the anhydrous charge carriers were immobilized on inorganic nanoparticles and incorporated into PEMs formulations. Nanoparticles with diameters ranging from 50-200 nm were synthesized via a sol-gel process and the desired anhydrous charge carriers immobilized on their surfaces. Nanocomposite PEMs were prepared using melt-processing technologies, by blending the grafted nanoparticles and fluorinated polymers such as poly (vinylidene fluoride) (PVDF) and ionomers such as Nafion. This paper presented the properties of the PEMs developed as a function of nanoparticles size and content, as well as the proton conductivity at controlled temperature and RH.

  11. Development of Melting Crucible Materials of Metallic Fuel Slug for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. H.; Lee, C. T.; Oh, S. J.; Kim, S. K.; Lee, C. B.; Ko, Y. M.; Woo, W. M

    2010-01-15

    The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr-Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized.

  12. Investigation of molybdate melts as an alternative method of reprocessing used nuclear fuel

    Science.gov (United States)

    Hames, Amber L.; Tkac, Peter; Paulenova, Alena; Willit, James L.; Williamson, Mark A.

    2017-04-01

    An investigation of molybdate melts containing sodium molybdate (Na2MoO4) and molybdenum trioxide (MoO3) to achieve the separation of uranium from fission products by crystallization has been performed. The separation is based on the difference in solubility of the fission product metal oxides compared to the uranium oxide or molybdate in the molybdate melt. The molybdate melt dissolves uranium dioxide at high temperatures, and upon cooling, uranium precipitates as uranium dioxide or molybdate, whereas the fission product metals remain soluble in the melt. Small-scale experiments using gram quantities of uranium dioxide have been performed to investigate the feasibility of UO2 purification from the fission products. The composition of the uranium precipitate as well as data for partitioning of several fission product surrogates between the uranium precipitate and molybdate melt for various melt compositions are presented and discussed. The fission products Cs, Sr, Ru and Rh all displayed very large distribution ratios. The fission products Zr, Pd, and the lanthanides also displayed good distribution ratios (D > 10). A melt consisting of 20 wt% MoO3- 50 wt% Na2MoO4-30 wt% UO2 heated to 1313 K and cooled to 1123 K for the physical separation of the UO2 product from the melt, and washed once with Na2MoO4 displays optimum conditions for separation of the UO2 from the fission products.

  13. Review of Severe Accident Phenomena in LWR and Related Severe Accident Analysis Codes

    Directory of Open Access Journals (Sweden)

    Muhammad Hashim

    2013-04-01

    Full Text Available Firstly, importance of severe accident provision is highlighted in view of Fukushima Daiichi accident. Then, extensive review of the past researches on severe accident phenomena in LWR is presented within this study. Various complexes, physicochemical and radiological phenomena take place during various stages of the severe accidents of Light Water Reactor (LWR plants. The review deals with progression of the severe accidents phenomena by dividing into core degradation phenomena in reactor vessel and post core melt phenomena in the containment. The development of various computer codes to analyze these severe accidents phenomena is also summarized in the review. Lastly, the need of international activity is stressed to assemble various severe accidents related knowledge systematically from research organs and compile them on the open knowledge base via the internet to be available worldwide.

  14. Icare/Cathare coupling: three-dimensional thermal hydraulics of severe LWR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Guillard, V.; Fichot, F. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, Dept. de Recherches en Securite, DRS, 92 (France); Boudier, P.; Parent, M. [CEA Grenoble, Dir. des Reacteurs Nucleaires, DRN, 38 (France); Roser, R. [Communication et Systemes Systemes d' Information, CS SI, 38 - Fontaine (France)

    2001-07-01

    In the phenomenology of severe LWR accidents considered in safety studies, the accidental sequences can be divided into three phases: the initial phase, where no severe damage of fuel or control rods and structures occurs; the early core degradation phase, where limited material melting and relocation takes place; and the late core degradation phase during which substantial material relocation happens, molten pools and debris beds can form and corium may fall into the lower plenum and, in case of vessel failure, come into the containment. The CATHARE2 code is a system code which has been developed by CEA for IPSN, EDF and FRAMATOME to describe the thermal-hydraulics behavior of a whole PWR circuit during the first of these three phases, with a core degradation model limited to clad rupture. The ICARE2 code, developed by IPSN, allows the complete description of early and late core degradation phases, with a thermal-hydraulics model limited to the vessel, initial and boundary conditions being provided by a system code. The aim of this paper is to present the main features of the new version of the coupling, ICARE/CATHARE V2. First, the general characteristics of ICARE2 V3mod1 and CATHARE2 V1.5 standard codes, dealing with physical models and numerical aspects, are described. Second, the technical features of the coupling between the two codes are detailed. At last, some results of ICARE/CATHARE V2 calculations are presented which demonstrate the ability of the code to simulate a severe accident in a PWR and notably to describe multi-dimensional effects occurring in the core during the LOCA and degradation phases. (authors)

  15. A view of treatment process of melted nuclear fuel on a severe accident plant using a molten salt system

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, R.; Takahashi, Y.; Nakamura, H.; Mizuguchi, K. [Power and Industrial Research and Development Center, Toshiba Corporation Power Systems Company, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan); Oomori, T. [Chemical System Design and Engineering Department, Toshiba Corporation Power Systems Company, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan)

    2013-07-01

    At severe accident such as Fukushima Daiichi Nuclear Power Plant Accident, the nuclear fuels in the reactor would melt and form debris which contains stable UO2-ZrO2 mixture corium and parts of vessel such as zircaloy and iron component. The requirements for solution of issues are below; -) the reasonable treatment process of the debris should be simple and in-situ in Fukushima Daiichi power plant, -) the desirable treatment process is to take out UO{sub 2} and PuO{sub 2} or metallic U and TRU metal, and dispose other fission products as high level radioactive waste; and -) the candidate of treatment process should generate the smallest secondary waste. Pyro-process has advantages to treat the debris because of the high solubility of the debris and its total process feasibility. Toshiba proposes a new pyro-process in molten salts using electrolysing Zr before debris fuel being treated.

  16. TREAT Neutronics Analysis of Water-Loop Concept Accommodating LWR 9-rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Hill, Connie M.; Woolstenhulme, Nicolas E.; Parry, James R.; Bess, John D.; Housley, Gregory K.

    2016-09-01

    Abstract. Simulation of a variety of transient conditions has been successfully achieved in the Transient Reactor Test (TREAT) facility during operation between 1959 and 1994 to support characterization and safety analysis of nuclear fuels and materials. A majority of previously conducted tests were focused on supporting sodium-cooled fast reactor (SFR) designs. Experiments evolved in complexity. Simulation of thermal-hydraulic conditions expected to be encountered by fuels and materials in a reactor environment was realized in the development of TREAT sodium loop experiment vehicles. These loops accommodated up to 7-pin fuel bundles and served to simulate more closely the reactor environment while safely delivering large quantities of energy into the test specimen. Some of the immediate TREAT restart operations will be focused on testing light water reactor (LWR) accident tolerant fuels (ATF). Similar to the sodium loop objectives, a water loop concept, developed and analyzed in the 1990’s, aimed at achieving thermal-hydraulic conditions encountered in commercial power reactors. The historic water loop concept has been analyzed in the context of a reactivity insertion accident (RIA) simulation for high burnup LWR 2-pin and 3-pin fuel bundles. Findings showed sufficient energy could be deposited into the specimens for evaluation. Similar results of experimental feasibility for the water loop concept (past and present) have recently been obtained using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. The old water loop concept required only two central TREAT core grid spaces. Preparation for future experiments has resulted in a modified water loop conceptual design designated the TREAT water environment recirculating loop (TWERL). The current TWERL design requires nine TREAT core grid spaces in order to place the water recirculating pump under the TREAT core. Due to the effectiveness of water moderation, neutronics analysis shows that removal of seven additional

  17. Anodic dissolution of irradiated metallic fuels in LiCl-KCl melt

    Science.gov (United States)

    Murakami, T.; Kato, T.; Rodrigues, A.; Ougier, M.; Iizuka, M.; Koyama, T.; Glatz, J.-P.

    2014-09-01

    Electrorefining is the main step in pyro-process of spent nuclear fuels, where actinides are recovered and separated from fission products. In the present study, electrorefining of irradiated metallic fuels called METAPHIX-1 (U-19 wt%Pu-10 wt%Zr alloy irradiated at PHENIX reactor, approximate maximum burn-up 2.5 at%) was performed. A major focus was on minimization of Zr co-dissolution from spent metallic fuels to reduce the burden to the pyro-process. Based on the ICP-MS analysis results and the SEM-EDX observations, the anodic dissolution behavior of the irradiated metallic fuels and the mass balances of actinides and fission products during the electrorefining were evaluated.

  18. Anodic dissolution of irradiated metallic fuels in LiCl–KCl melt

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T., E-mail: m-tsuyo@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (CRIEPI), Komaeshi, Tokyo 201-8511 (Japan); Kato, T. [Central Research Institute of Electric Power Industry (CRIEPI), Komaeshi, Tokyo 201-8511 (Japan); Rodrigues, A.; Ougier, M. [Joint Research Center–Institute for Transuranium Elements (JRC–ITU), P.O. Box 2340, 76125 Karlsruhe (Germany); Iizuka, M.; Koyama, T. [Central Research Institute of Electric Power Industry (CRIEPI), Komaeshi, Tokyo 201-8511 (Japan); Glatz, J.-P. [Joint Research Center–Institute for Transuranium Elements (JRC–ITU), P.O. Box 2340, 76125 Karlsruhe (Germany)

    2014-09-15

    Electrorefining is the main step in pyro-process of spent nuclear fuels, where actinides are recovered and separated from fission products. In the present study, electrorefining of irradiated metallic fuels called METAPHIX-1 (U–19 wt%Pu–10 wt%Zr alloy irradiated at PHENIX reactor, approximate maximum burn-up 2.5 at%) was performed. A major focus was on minimization of Zr co-dissolution from spent metallic fuels to reduce the burden to the pyro-process. Based on the ICP-MS analysis results and the SEM–EDX observations, the anodic dissolution behavior of the irradiated metallic fuels and the mass balances of actinides and fission products during the electrorefining were evaluated.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    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

  20. In-core materials testing under LWR conditions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  1. Combustion performance of an aluminum melting furnace operating with liquid fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nieckele, Angela Ourivio; Naccache, Monica Feijo; Gomes, Marcos Sebastiao de P. [Pontificia Universidade Catolica (PUC-Rio), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Mecanica], E-mails: nieckele@puc-rio.br, naccache@puc-rio.br, mspgomes@puc-rio.br

    2010-10-15

    The characteristics associated with the delivery of the fuel to be used as the energy source in any industrial combustion equipment are of extreme importance, as for example, in improving the performance of the combustion process and in the preservation of the equipment. A clean and efficient combustion may be achieved by carefully selecting the fuel and oxidant, as well as the operational conditions of the delivery system for both. In the present work, numerical simulations were carried out using the commercial code FLUENT for analyzing some of the relevant operational conditions inside an aluminum reverb furnace employing liquid fuel and air as the oxidant. Different fuel droplets sizes as well as inlet droplet stream configurations were examined. These characteristics, associated with the burner geometry and the fuel dispersion and delivery system may affect the flame shape, and consequently the temperature and the heat flux distribution within the furnace. Among the results obtained in the simulations, it was shown the possible damages to the equipment, which may occur as a result of the combustion process, if the flame is too long or too intense and concentrated. (author)

  2. Corrigendum to "Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions" [J. Nucl. Mater. 448 (2014) 520-533

    Science.gov (United States)

    Ott, L. J.; Robb, K. R.; Wang, D.

    2015-06-01

    In Section 5.2, certain material properties for "FeCrAl oxide" were not modeled based on "stainless steel oxide" as indicated in the text. Instead, the "FeCrAl oxide" material properties were modeled using the default properties in MELCOR for "zirconium oxide". The properties affected are the FeCrAl oxide density, specific heat, enthalpy, thermal conductivity, melting point, and latent heat of fusion. Table 5.1 and Figs. 5.1a-d from Section 5.2 have been corrected below. As discussed below, the overall conclusions of the paper remain unchanged.

  3. Multilevel transport solution of LWR reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Jose Ignacio Marquez Damian; Cassiano R.E. de Oliveira; HyeonKae Park

    2008-09-01

    This work presents a multilevel approach for the solution of the transport equation in typical LWR assemblies and core configurations. It is based on the second-order, even-parity formulation of the transport equation, which is solved within the framework provided by the finite element-spherical harmonics code EVENT. The performance of the new solver has been compared with that of the standard conjugate gradient solver for diffusion and transport problems on structured and unstruc-tured grids. Numerical results demonstrate the potential of the multilevel scheme for realistic reactor calculations.

  4. On the rational formulation of alternative fuels: melting point and net heat of combustion predictions for fuel compounds using machine learning methods.

    Science.gov (United States)

    Saldana, D A; Starck, L; Mougin, P; Rousseau, B; Creton, B

    2013-01-01

    We report the development of predictive models for two fuel specifications: melting points (T(m)) and net heat of combustion (Δ(c)H). Compounds inside the scope of these models are those likely to be found in alternative fuels, i.e. hydrocarbons, alcohols and esters. Experimental T(m) and Δ(c)H values for these types of molecules have been gathered to generate a unique database. Various quantitative structure-property relationship (QSPR) approaches have been used to build models, ranging from methods leading to multi-linear models such as genetic function approximation (GFA), or partial least squares (PLS) to those leading to non-linear models such as feed-forward artificial neural networks (FFANN), general regression neural networks (GRNN), support vector machines (SVM), or graph machines. Except for the case of the graph machines method for which the only inputs are SMILES formulae, previously listed approaches working on molecular descriptors and functional group count descriptors were used to develop specific models for T(m) and Δ(c)H. For each property, the predictive models return slightly different responses for each molecular structure. Therefore, models labelled as 'consensus models' were built by averaging values computed with selected individual models. Predicted results were then compared with experimental data and with predictions of models in the literature.

  5. MATPRO: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P.E.; Thompson, L.B. (eds.)

    1976-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Documentation and formulations that are generally semiempirical in nature are presented for uranium dioxide and mixed uranium-plutonium dioxide fuel, zircaloy cladding, gas mixture, and LWR fuel rod material properties.

  6. Survey of LWR environmental control technology performance and cost

    Energy Technology Data Exchange (ETDEWEB)

    Heeb, C.M.; Aaberg, R.L.; Cole, B.M.; Engel, R.L.; Kennedy, W.E. Jr.; Lewallen, M.A.

    1980-03-01

    This study attempts to establish a ranking for species that are routinely released to the environment for a projected nuclear power growth scenario. Unlike comparisons made to existing standards, which are subject to frequent revision, the ranking of releases can be used to form a more logical basis for identifying the areas where further development of control technology could be required. This report describes projections of releases for several fuel cycle scenarios, identifies areas where alternative control technologies may be implemented, and discusses the available alternative control technologies. The release factors were used in a computer code system called ENFORM, which calculates the annual release of any species from any part of the LWR nuclear fuel cycle given a projection of installed nuclear generation capacity. This survey of fuel cycle releases was performed for three reprocessing scenarios (stowaway, reprocessing without recycle of Pu and reprocessing with full recycle of U and Pu) for a 100-year period beginning in 1977. The radioactivity releases were ranked on the basis of a relative ranking factor. The relative ranking factor is based on the 100-year summation of the 50-year population dose commitment from an annual release of radioactive effluents. The nonradioactive releases were ranked on the basis of dilution factor. The twenty highest ranking radioactive releases were identified and each of these was analyzed in terms of the basis for calculating the release and a description of the currently employed control method. Alternative control technology is then discussed, along with the available capital and operating cost figures for alternative control methods.

  7. Closed Fuel Cycle Waste Treatment Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, J. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Collins, E. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Crum, J. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, S. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Garn, T. G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gombert, D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maio, V. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Matyas, J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Nenoff, T. M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Riley, B. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sevigny, G. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thallapally, P. K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, J. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-02-01

    with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

  8. A Brief Review of Past INL Work Assessing Radionuclide Content in TMI-2 Melted Fuel Debris: The Use of 144Ce as a Surrogate for Pu Accountancy

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Chichester; S. J. Thompson

    2013-09-01

    This report serves as a literature review of prior work performed at Idaho National Laboratory, and its predecessor organizations Idaho National Engineering Laboratory (INEL) and Idaho National Engineering and Environmental Laboratory (INEEL), studying radionuclide partitioning within the melted fuel debris of the reactor of the Three Mile Island 2 (TMI-2) nuclear power plant. The purpose of this review is to document prior published work that provides supporting evidence of the utility of using 144Ce as a surrogate for plutonium within melted fuel debris. When the TMI-2 accident occurred no quantitative nondestructive analysis (NDA) techniques existed that could assay plutonium in the unconventional wastes from the reactor. However, unpublished work performed at INL by D. W. Akers in the late 1980s through the 1990s demonstrated that passive gamma-ray spectrometry of 144Ce could potentially be used to develop a semi-quantitative correlation for estimating plutonium content in these materials. The fate and transport of radioisotopes in fuel from different regions of the core, including uranium, fission products, and actinides, appear to be well characterized based on the maximum temperature reached by fuel in different parts of the core and the melting point, boiling point, and volatility of those radioisotopes. Also, the chemical interactions between fuel, fuel cladding, control elements, and core structural components appears to have played a large role in determining when and how fuel relocation occurred in the core; perhaps the most important of these reaction appears to be related to the formation of mixed-material alloys, eutectics, in the fuel cladding. Because of its high melting point, low volatility, and similar chemical behavior to plutonium, the element cerium appears to have behaved similarly to plutonium during the evolution of the TMI-2 accident. Anecdotal evidence extrapolated from open-source literature strengthens this logical feasibility for

  9. Oxidation Protection of Uranium Nitride Fuel using Liquid Phase Sintering

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Paul A. Lessing

    2012-03-01

    Two methods are proposed to increase the oxidation resistance of uranium nitride (UN) nuclear fuel. These paths are: (1) Addition of USi{sub x} (e.g. U3Si2) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with various compounds (followed by densification via Spark Plasma Sintering or Liquid Phase Sintering) that will greatly increase oxidation resistance. The advantages (high thermal conductivity, very high melting point, and high density) of nitride fuel have long been recognized. The sodium cooled BR-10 reactor in Russia operated for 18 years on uranium nitride fuel (UN was used as the driver fuel for two core loads). However, the potential advantages (large power up-grade, increased cycle lengths, possible high burn-ups) as a Light Water Reactor (LWR) fuel are offset by uranium nitride's extremely low oxidation resistance (UN powders oxidize in air and UN pellets decompose in hot water). Innovative research is proposed to solve this problem and thereby provide an accident tolerant LWR fuel that would resist water leaks and high temperature steam oxidation/spalling during an accident. It is proposed that we investigate two methods to increase the oxidation resistance of UN: (1) Addition of USi{sub x} (e.g. U{sub 3}Si{sub 2}) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with compounds (followed by densification via Spark Plasma Sintering) that will greatly increase oxidation resistance.

  10. Microstructural Characterization and Wear Behavior of Nano-Boride Dispersed Coating on AISI 304 Stainless Steel by Hybrid High Velocity Oxy-Fuel Spraying Laser Surface Melting

    Science.gov (United States)

    Sharma, Prashant; Majumdar, Jyotsna Dutta

    2015-07-01

    The current study concerns the detailed microstructural characterization and investigation of wear behavior of nano-boride dispersed coating developed on AISI 304 stainless steel by high velocity oxy-fuel spray deposition of nickel-based alloy and subsequent laser melting. There is a significant refinement and homogenization of microstructure with improvement in microhardness due to laser surface melting (1200 VHN as compared to 945 VHN of as-sprayed and 250 VHN of as-received substrate). The high temperature phase stability of the as-coated and laser melted surface has been studied by differential scanning calorimeter followed by detailed phase analysis at room and elevated temperature. There is a significant improvement in wear resistance of laser melted surface as compared to as-sprayed and the as-received one due to increased hardness and reduced coefficient of friction. The mechanism of wear has been investigated in details. Corrosion resistance of the coating in a 3.56 wt pct NaCl solution is significantly improved (4.43 E-2 mm/year as compared to 5 E-1 mm/year of as-sprayed and 1.66 mm/year of as-received substrate) due to laser surface melting as compared to as-sprayed surface.

  11. Study of the potential uses of the Barnwell Nuclear Fuel Plant (BNFP). Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-25

    The purpose of this study is to provide an evaluation of possible international and domestic uses for the Barnwell Nuclear Fuel Plant, located in South Carolina, at the conclusion of the International Nuclear Fuel Cycle Evaluation. Four generic categories of use options for the Barnwell plant have been considered: storage of spent LWR fuel; reprocessing of LWR spent fuel; safeguards development and training; and non-use. Chapters are devoted to institutional options and integrated institutional-use options.

  12. Preliminary concepts: coordinated safeguards for materials management in a thorium--uranium fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Hakkila, E.A.; Barnes, J.W.; Dayem, H.A.; Dietz, R.J.; Shipley, J.P.

    1978-10-01

    This report addresses preliminary concepts for coordinated safeguards materials management in a typical generic thorium--uranium-fueled light-water reactor (LWR) fuels reprocessing plant. The reference facility is designed to recover thorium and uranium from first-generation (denatured /sup 235/U) startup fuels, first-recycle and equilibrium (denatured /sup 233/U) thorium--uranium LWR fuels, and to recover the plutonium generated in the /sup 238/U denaturant as well. 12 figures, 3 tables.

  13. Analysis of pellet cladding interaction and creep of U 3SIi2 fuel for use in light water reactors

    Science.gov (United States)

    Metzger, Kathryn E.

    Following the accident at the Fukushima plant, enhancing the accident tolerance of the light water reactor (LWR) fleet became a topic of serious discussion. Under the direction of congress, the DOE office of Nuclear Energy added accident tolerant fuel development as a primary component to the existing Advanced Fuels Program. The DOE defines accident tolerant fuels as fuels that "in comparison with the standard UO2- Zircaloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events." To be economically viable, proposed accident tolerant fuels and claddings should be backward compatible with LWR designs, provide significant operating cost improvements such as power uprates, increased fuel burnup, or increased cycle length. In terms of safety, an alternative fuel pellet must have resistance to water corrosion comparable to UO2, thermal conductivity equal to or larger than that of UO2, and a melting temperature that allows the material to remain solid under power reactor conditions. Among the candidates, U3Si2 has a number of advantageous thermophysical properties, including; high density, high thermal conductivity at room temperature, and a high melting temperature. These properties support its use as an accident tolerant fuel while its high uranium density is capable of supporting uprates to the LWR fleet. This research characterizes U3Si2 pellets and analyzes U3Si2 under light water reactor conditions using the fuel performance code BISON. While some thermophysical properties for U3Si2 have been found in the literature, the irradiation behavior is sparse and limited to experience with dispersion fuels. Accordingly, the creep behavior for U3Si2 has been unknown, making it

  14. Corrosion report for the U-Mo fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Bennett, Wendy D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Fuller, E. S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hardy, John S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2014-08-28

    The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30°C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

  15. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  16. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  17. Integrated Computational Modeling of Water Side Corrosion in Zirconium Metal Clad Under Nominal LWR Operating Conditions

    Science.gov (United States)

    Aryanfar, Asghar; Thomas, John; Van der Ven, Anton; Xu, Donghua; Youssef, Mostafa; Yang, Jing; Yildiz, Bilge; Marian, Jaime

    2016-10-01

    A mesoscopic chemical reaction kinetics model to predict the formation of zirconium oxide and hydride accumulation light-water reactor (LWR) fuel clad is presented. The model is designed to include thermodynamic information from ab initio electronic structure methods as well as parametric information in terms of diffusion coefficients, thermal conductivities and reaction constants. In contrast to approaches where the experimentally observed time exponents are captured by the models by design, our approach is designed to be predictive and to provide an improved understanding of the corrosion process. We calculate the time evolution of the oxide/metal interface and evaluate the order of the chemical reactions that are conducive to a t 1/3 dependence. We also show calculations of hydrogen cluster accumulation as a function of temperature and depth using spatially dependent cluster dynamics. Strategies to further cohesively integrate the different elements of the model are provided.

  18. Integrated Computational Modeling of Water Side Corrosion in Zirconium Metal Clad Under Nominal LWR Operating Conditions

    Science.gov (United States)

    Aryanfar, Asghar; Thomas, John; Van der Ven, Anton; Xu, Donghua; Youssef, Mostafa; Yang, Jing; Yildiz, Bilge; Marian, Jaime

    2016-11-01

    A mesoscopic chemical reaction kinetics model to predict the formation of zirconium oxide and hydride accumulation light-water reactor (LWR) fuel clad is presented. The model is designed to include thermodynamic information from ab initio electronic structure methods as well as parametric information in terms of diffusion coefficients, thermal conductivities and reaction constants. In contrast to approaches where the experimentally observed time exponents are captured by the models by design, our approach is designed to be predictive and to provide an improved understanding of the corrosion process. We calculate the time evolution of the oxide/metal interface and evaluate the order of the chemical reactions that are conducive to a t 1/3 dependence. We also show calculations of hydrogen cluster accumulation as a function of temperature and depth using spatially dependent cluster dynamics. Strategies to further cohesively integrate the different elements of the model are provided.

  19. Melt crystallization for refinement of triolein and palmitic acid mixture as a model waste oil for biodiesel fuel production

    Science.gov (United States)

    Fukui, Keisuke; Maeda, Kouji; Kuramochi, Hidetoshi

    2013-06-01

    Melt crystallization using an annular vessel with two circular cylinders was applied to produce high-quality vegetable oil from waste oil. The inner cylinder was cooled at a constant rate and rotated, and the outer cylinder was heated at a constant temperature. The melt was solidified on the inner cylinder surface. The binary system of triolein and palmitic acid was used as the model waste oil. We measured the distribution coefficient of triolein. Suitable operation conditions were proposed to attain a high yield and a high purity of triolein from waste oil. The distribution coefficient correlated well with the theoretical equation derived on the basis of the "local lever rule" at the interface of the crystal layer and melt [1].

  20. In-can melting process and equipment development from 1974 to 1978

    Energy Technology Data Exchange (ETDEWEB)

    Blair, H. Thomas

    1979-08-01

    Both the defense HLLW stores in tanks presently and the HLLW from proposed reprocessing of commercial LWR fuel can be vitrified as borosilicate glass in containers made of 300-series stainless steel by the ICM (in-can melting) process. Melting rates of 50 kg/h in 12-in.-dia cans and 117 kg/h in 28-in.-dia cans can be achieved in the ICM by using the rising-level charging method and internal heat-transfer plate assemblies in the cans. The ICM process can be monitored and remotely controlled without the aid of instrumentation attached to the waste can. The ICM process is compatible with both heated-wall spray calciners and fluidized-bed calciners. The ICM process causes residual tensile stresses as high as the yield strength in vitrified product containers made of 300-series stainless steel. Spall due to oxidation of the exterior of the can during an ICM process can be prevented by using an inert cover gas, by putting a protective coating on the can surface, or by using an oxidation-resistant alloy. Processing problems are minimized and product quality is improved when the complete can is located inside the furnace chamber by setting it on the hearth. A maximum of 24 kW and an average of 15 kW is required per 15-in.-high furnace zone to melt waste borosilicate glass at a rate of 117 kg/h in a 28-in.-dia ICM.

  1. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Science.gov (United States)

    Flaspoehler, Timothy; Petrovic, Bojan

    2016-02-01

    One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV). While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR) Power Plant, including DPA (displacements per atom) and fast neutron fluence (>1 MeV and >0.1MeV). I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling) methodology to bias a fixed-source MC (Monte Carlo) simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  2. Radiation Damage Assessment in the Reactor Pressure Vessel of the Integral Inherently Safe Light Water Reactor (I2S-LWR

    Directory of Open Access Journals (Sweden)

    Flaspoehler Timothy

    2016-01-01

    Full Text Available One of the major limiting factors to nuclear reactors lifetime is the radiation-induced material damage in the Reactor Pressure Vessel (RPV. While older reactors were designed assuming a 40-year operating lifetime, new reactor designs are expected to have lifetimes up to 100 years. For safe operation, the integrity of the RPV must be ensured against significant material property changes. In this work, typical neutron damage indicators are calculated in the RPV of the I2S-LWR (Integral Inherently Safe LWR Power Plant, including DPA (displacements per atom and fast neutron fluence (>1 MeV and >0.1MeV. I2S-LWR is a PWR of integral design, which means that its wider downcomer provides additional shielding to the vessel. However, its higher core power density and longer lifetime may offset this advantage. In order to accurately represent the neutron environment for RPV damage assessment, a detailed model based on the preliminary design specifications of the I2S-LWR was developed to be used in the MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations sequence of the Scale6.1 code package. MAVRIC uses the CADIS (Consistent Adjoint-Driven Importance Sampling methodology to bias a fixed-source MC (Monte Carlo simulation. To establish the upper limit of a bounding envelope, a flat-source distribution was used. For the low limit, a center-peaked source was generated using the KENO-VI criticality sequence assuming uniform fresh fuel core. Results based on the preliminary I2S-LWR model show that DPA rates and fast fluence rates are conservatively 75% lower than in typical PWRs being operated currently in the US.

  3. Improving the safety of LWR power plants. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-01

    This report documents the results of the Study to identify current, potential research issues and efforts for improving the safety of Light Water Reactor (LWR) power plants. This final report describes the work accomplished, the results obtained, the problem areas, and the recommended solutions. Specifically, for each of the issues identified in this report for improving the safety of LWR power plants, a description is provided in detail of the safety significance, the current status (including information sources, status of technical knowledge, problem solution and current activities), and the suggestions for further research and development. Further, the issues are ranked for action into high, medium, and low priority with respect to primarily (a) improved safety (e.g. potential reduction in public risk and occupational exposure), and secondly (b) reduction in safety-related costs (improving or maintaining level of safety with simpler systems or in a more cost-effective manner).

  4. Understanding EUV mask blank surface roughness induced LWR and associated roughness requirement

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Pei-Yang [Intel Corp., Santa Clara, CA (United States); Zhang, Guojing [Intel Corp., Santa Clara, CA (United States); Gullickson, Eric M. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Goldberg, Kenneth A. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Benk, Markus P. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2015-03-01

    Extreme ultraviolet lithography (EUVL) mask multi-layer (ML) blank surface roughness specification historically comes from blank defect inspection tool requirement. Later, new concerns on ML surface roughness induced wafer pattern line width roughness (LWR) arise. In this paper, we have studied wafer level pattern LWR as a function of EUVL mask surface roughness via High-NA Actinic Reticle Review Tool. We found that the blank surface roughness induced LWR at current blank roughness level is in the order of 0.5nm 3σ for NA=0.42 at the best focus. At defocus of ±40nm, the corresponding LWR will be 0.2nm higher. Further reducing EUVL mask blank surface roughness will increase the blank cost with limited benefit in improving the pattern LWR, provided that the intrinsic resist LWR is in the order of 1nm and above.

  5. Assessment of LWR piping design loading based on plant operating experience

    Energy Technology Data Exchange (ETDEWEB)

    Svensson, P. O.

    1980-08-01

    The objective of this study has been to: (1) identify current Light Water Reactor (LWR) piping design load parameters, (2) identify significant actual LWR piping loads from plant operating experience, (3) perform a comparison of these two sets of data and determine the significance of any differences, and (4) make an evaluation of the load representation in current LWR piping design practice, in view of plant operating experience with respect to piping behavior and response to loading.

  6. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  7. North Korean nuclear issues and the LWR project; analysis of the key technologies

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Koo; Kwack, E. H.; Shin, J. S.; An, J. S.; Lee, J. U.; Kim, H. T.; Kim, J. S.; Yoon, Y. C

    2000-11-01

    Year 2000 will be remembered as an epoch making period between two Koreas. Korean nuclear industries became activated with the KEDO LWR main contracts entering into force in February, 2000. Respective design, manufacturing and construction activities are mobilized in accordance with the total project schedule of about 100 months. What started out as the nuclear power plant design standardization project in the early '80s, is now being implemented as repeat construction of KSNPs at Yonggwang and Ulchin sites as well as at Kumho site in the DPRK. However, the KEDO construction schedule and the past nuclear inconsistency issues are closely linked due to the nonproliferation concerns. In practice, the IAEA must come to the conclusion that the correctness and completeness must be fulfilled before delivery of the first key component for the KEDO LWR unit 1. While the IAEA verification process tends to focus on the nuclear materials accountancy control, longer term objective between two Koreas is bound to take the form of confidence building. It is necessary to analyse the nuclear research and production facilities in order to make proper evaluation of a nation's nuclear capabilities. Close assessment on development status of graphite moderated reactors and their operation history, spent fuel reprocessing facilities, and HEU production capabilities would be essential. In addition, illicit trafficking possibilities should be addressed. Chapter 1 describes the graphite moderated reactors in general; Chapter 2 describes various reprocessing processes and their detection capabilities; Chapter 3 contains possible uranium enrichment processes with their detection capabilities, and Chapter 4 summarizes the international treaties in illicit trafficking control with the IAEA database.

  8. Investigation on Melt-Structure-Water Interactions (MSWI) during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Yang, Z.L.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Haraldsson, H.O.; Li, H.X.; Konovakhin, M.; Paladino, D.; Leung, W.H [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1999-08-01

    This report is the final report for the work performed in 1998 in the research project Melt Structure Water Interactions (MSWI), under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The present report describes results of advanced analytical and experimental studies concerning melt-water-structure interactions during the course of a hypothetical severe core meltdown accident in a light water reactor (LWR). Emphasis has been placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. Many of the investigations performed in support of this project have produced papers which have been published in the proceedings of technical meetings. A short summary of the results achieved in these papers is provided in this overview. Both experimental and analytical studies were performed to improve knowledge about phenomena of melt-structure-water interactions. We believe that significant technical advances have been achieved during the course of these studies. It was found that: the solidification has a strong effect on the drop deformation and breakup. Initially appearing at the drop surface and, later, thickening inwards, the solid crust layer dampens the instability waves on the drop surface and, therefore, hinders drop deformation and breakup. The drop thermal properties also affect the thermal behavior of the drop and, therefore, have impact on its deformation behavior. The jet fragmentation process is a function of many related phenomena. The fragmentation rate depends not only on the traditional parameters, e.g. the Weber number, but also on the melt physical properties, which change as the melt cools down from the liquidus to the solidus temperature. Additionally, the crust formed on the surface of the melt jet will also reduce the propensity

  9. Characteriztion of particulate plutonium released in fuel cycle operations

    Energy Technology Data Exchange (ETDEWEB)

    Seefeldt, W.B.; Mecham, W.J.; Steindler, M.J.

    1976-05-01

    An estimate of the plutonium source terms is made for the fuel cycles of three reactor types on the basis of currently applied, currently available, and estimated future technology. The three reactors are LWR-U, LWR-Pu, and LMFBR. The source terms are characterized as to quantity, form, and particle size distribution. Historical operating data for existing plants and the state of the art of the technology of air cleaning are reviewed.

  10. Radiation protection of MOX-fuel by doping with Pa-23 and U-232

    NARCIS (Netherlands)

    Kryuchkov, EF; Glebov, VB; Apse, VA; Shmelev, AN

    2005-01-01

    The paper addresses the problem of MOX-fuel self-protection at the "Spent Fuel Standard" level and more during full cycle of MOX-fuel management. Under conditions of the closed LWR cycle the proliferation-resistance levels were evaluated for fresh and spent MOX-fuel doped with Pa-231 and U-232. Acco

  11. Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors; Validacion del codigo AZTRAN 1.1 con problemas Benchmark de reactores LWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M., E-mail: amhed.jvq@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S{sub N}, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO{sub 2} cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

  12. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  13. Reprocessing method for spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hoshikawa, Tadahiro; Sawa, Toshio; Suzuoki, Akira [Hitachi Ltd., Tokyo (Japan); Takashima, Yoichi; Kumagai, Mikiro

    1998-09-29

    The present invention provides a method of reprocessing spent fuels to form MOX having a Pu/U ratio suitable to fuels of LWR or fast reactors and uranium oxides of fuels of an LWR reactor. In a brief separation step for uranium, carbonate is added to a nitric acid solution in which spent fuels are dissolved, to dissolve a portion of uranium in the nitric acid solution. The residual uranium, plutonium and fission products are made into complexes of carboxylic acid ions and precipitated. The precipitated complexes of carboxylic acid ions are brought into contact with a different nitric acid solution to recover the uranium, plutonium and fission products. The concentration of the carbonate in the nitric acid solution in which uranium is partially dissolved is determined in accordance with the plutonium/uranium ratio based on the relation between the saturation concentration of uranium to the concentration of carbonate in the nitric acid solution. (T.M.)

  14. Investigation of valve failure problems in LWR power plants

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).

  15. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, James [Univ. of Wisconsin, Madison, WI (United States); Butt, Darryl [Boise State Univ., ID (United States); Meyer, Mitchell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Corporation, Pittsburgh, PA (United States)

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  16. Thermodynamic analysis for high burn-up fuel internal chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Fuji, Kensho; Kyoh, Bunkei [Kinki Univ., Higashi-Osaka, Osaka (Japan). Faculty of Science and Technology

    1997-09-01

    Chemical states of fission products and actinide elements in high burn-up LWR fuel pellets have been analyzed thermodynamically using the computer program SOLGASMIX-PV. Calculations with this computer code have been performed for a complex multi-component system, which comprises 54 chemical species. The analysis shows that neither alkali nor alkaline-earth uranates are formed, but alkali and alkaline-earth molybdates exist in irradiated LWR fuel pellets in contrast with their post irradiation examinations. These molybdates tend to increase with increasing oxygen potential in the fuel under operating conditions, whereas the zirconates decrease. (author)

  17. A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

    Directory of Open Access Journals (Sweden)

    Jorge Pérez Mañes

    2014-01-01

    Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

  18. Fuel Thermal Expansion (FTHEXP). [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Reymann, G. A.

    1978-07-01

    A model is presented which deals with dimensional changes in LWR fuel pellets caused by changes in temperature. It is capable of dealing with any combination of UO/sub 2/ and PuO/sub 2/ in solid, liquid or mixed phase states, and includes expansion due to the solid-liquid phase change. The function FTHEXP models fuel thermal expansion as a function of temperature, fraction of PuO/sub 2/, and the fraction of fuel which is molten.

  19. Advanced Fuels Campaign FY 2014 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase; W. Edgar May

    2014-10-01

    The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance.

  20. FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ``Alternative Teams,`` chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S&S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT`s analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option.

  1. LWR and defectivity improvement on EUV track system

    Science.gov (United States)

    Harumoto, Masahiko; Stokes, Harold; Thouroude, Yan; Kaneyama, Koji; Pieczulewski, Charles; Asai, Masaya

    2016-03-01

    EUV lithography (EUVL) is well known to be a strong candidate for next generation, single exposure sub-30nm halfpitch lithography.[1] Furthermore, high-NA EUV exposure tool(s) released two years ago gave a strong impression by finer pattern results. On the other hand, it seems that the coat-develop track process remains very similar and in many aspects returns to KrF or ArF dry process fundamentals, but in practice a 26-32nm pitch patterning coat develop track process also has challenges with EUV resists. As access to EUV lithography exposures has become more readily available over the last five (5) years, several challenges and accomplishments in the track process have been reported, such as the improvement of ultra-thin film coating, CD uniformity, defectivity, line width roughness (LWR), and so on.[2-8] The coat-develop track process has evolved along with novel materials and metrology capability. Line width roughness (LWR) control and defect reduction are demonstrated utilizing the SOKUDO DUO coat-develop track system with ASML NXE:3100 and NXE:3300 exposures in the IMEC (Leuven, Belgium) cleanroom environment. Additionally, we will show the latest lithographic results obtained by novel processing approaches in the EUV coat develop track system.

  2. Reactor Physics Assessment of Thick Silicon Carbide Clad PWR Fuels

    Science.gov (United States)

    2013-06-01

    Loss of Coolant Accident LWR Light Water Reactor MOX Mixed Oxide Fuel MTC Moderator Temperature Coefficient MWd/kgIHM Megawatt days per...working only with UO2 and UO2/PuO2 mixed oxide ( MOX ) fuels. 3.1 Studsvik Core Management Software CASMO-4E and SIMULATE-3 are the primary computational

  3. Survey of nuclear fuel cycle economics: 1970--1985

    Energy Technology Data Exchange (ETDEWEB)

    Prince, B. E.; Peerenboom, J. P.; Delene, J. G.

    1977-03-01

    This report is intended to provide a coherent view of the diversity of factors that may affect nuclear fuel cycle economics through about 1985. The nuclear fuel cycle was surveyed as to past trends, current problems, and future considerations. Unit costs were projected for each step in the fuel cycle. Nuclear fuel accounting procedures were reviewed; methods of calculating fuel costs were examined; and application was made to Light Water Reactors (LWR) over the next decade. A method conforming to Federal Power Commission accounting procedures and used by utilities to account for backend fuel-cycle costs was described which assigns a zero net salvage value to discharged fuel. LWR fuel cycle costs of from 4 to 6 mills/kWhr (1976 dollars) were estimated for 1985. These are expected to reach 6 to 9 mills/kWr if the effect of inflation is included.

  4. FUEL SERVICES: Customer focused on Product Support during the whole Life Time

    Energy Technology Data Exchange (ETDEWEB)

    Langenberger, J.; Hummel, W.

    2015-07-01

    For more than 40 years, Fuel Services of AREVA has been delivering devices and providing on-site services primary at LWR worldwide. We support our worldwide customers in achieving safe and economic operation of the fuel assemblies (FA) and core components (CC) and have received excellent feedback from them. But the Fuel Services support goes beyond on-site activities. (Author)

  5. Experimental Results on Pouring and Underwater Liquid Melt Spreading and Energetic Melt-coolant Interaction

    OpenAIRE

    Konovalenko, Alexander; Karbojian, Aram; Kudinov, Pavel

    2012-01-01

    In a hypothetical light water reactor (LWR) core-melt accident with corium release from the reactor  vessel,  the  ultimate  containment  integrity  is  contingent  on  coolability  of  the decay-heated core debris. Pouring of melt into a pool of water located in the reactor cavity is considered in several designs of existing and new LWRs  as a part of severe accident (SA) management strategies. At certain conditions of melt release into the pool (e.g. large ratio of the  vessel  breach  size...

  6. Fission product releases at severe LWR accident conditions: ORNL/CEA measurements versus calculations

    Energy Technology Data Exchange (ETDEWEB)

    Andre, B.; Ducros, G.; Leveque, J.P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Osborne, M.F.; Lorenz, R.A. [Oak Ridge National Lab., TN (United States); Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Dept. de Protection de l`Environnement et des Installations

    1995-12-31

    Experimental programs in the United States and France have followed similar paths in supplying much of the data needed to analyze severe accidents. Both the HI/VI program, conducted at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (NRC), and the HEVA/VERCORS program, supported by IPSN-Commissariat a l`Energie Atomique (CEA) and carried out at the Centre d`Etudes Nucleaires de Grenoble, have studied fission product release from light water reactor (LWR) fuel samples during test sequences representative of severe accidents. Recognizing that more accurate data, i.e., a better defined source term, could reduce the safety margins included in the rather conservative source terms originating from WASH-1400, the primary objective of these programs has been to improve the data base concerning fission product release and behavior at high temperatures. To facilitate the comparison, a model based on fission product diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model is compared with the ORNL experimental data in a blind test. The two experimental programs used similar techniques in out-of-pile studies. Highly irradiated fuel samples were heated in radiofrequency induction furnaces to very high temperatures (up to 2700 K at ORNL and 2750 K at CEA) in oxidizing (H{sub 2}O), reducing (H{sub 2}) or mixed (H{sub 2}O+H{sub 2}) environments. The experimental parameters, which were chosen from calculated accident scenarios, did not duplicate specific accidents, but rather emphasized careful control of test conditions to facilitate extrapolation of the results to a wide variety of accident situations. This paper presents a broad and consistent database from ORNL and CEA release results obtained independently since the early 1980`S. A comparison of CORSOR and CORSOR Booth calculations, currently used in safety analysis, and the experimental results is presented and

  7. Neutronic experiment planning for the Fuels Refabrication and Development Program

    Energy Technology Data Exchange (ETDEWEB)

    Gore, B.F.; NcNeese, J.P.; Zimmerman, M.G.; Konzek, G.J.

    1979-12-01

    A program of experiments using /sup 233/UO/sub 2/ - ThO/sub 2/ fuel was proposed to provide new and improved neutronic and criticality data for thorium based nuclear fuels, in order to support the licensing of /sup 233/UO/sub 2/ - ThO/sub 2/ fuels in LWR cores. This would support the goal to develop technology for proliferation resistant fuel cycles to a point where fuel cycle choice is not limited by refabrication technology. The proposed experimental program is described in this document, along with initial planning and fuel acquisition activities undertaken during FY 1979. The program was terminated following notification that the DOE-sponsored denatured LWR Fuel Program which the experiments supported was to be discontinued.

  8. Metholology for the selection of LWR safety R and D projects. Phase I, status report

    Energy Technology Data Exchange (ETDEWEB)

    El-Sheikh, K. A.

    1980-03-01

    The objective of the LWR R and D Selection Methodology Program is to develop and demonstrate an R and D selection methodology appropriate for LWR safety technology. This report documents the development work from the program beginning in April, 1979 to the end of Fiscal Year 1979. The scope of work for this period included three tasks; methodology review (Task 1), measures development (Task 2), and methodology development for the first phase of application (Task 3). The accomplishments of these tasks are presented.

  9. Advanced Fuels Campaign FY 2015 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Braase, Lori Ann [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carmack, William Jonathan [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-29

    The mission of the Advanced Fuels Campaign (AFC) is to perform research, development, and demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This report is a compilation of technical accomplishment summaries for FY-15. Emphasis is on advanced accident-tolerant LWR fuel systems, advanced transmutation fuels technologies, and capability development.

  10. Release of segregated nuclides from spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, L.H.; Tait, J.C. [Atomic Energy Canada Ltd., Pinawa, MB (Canada). Whiteshell Laboratories

    1997-10-01

    The potential release of fission and activation products from spent nuclear fuel into groundwater after container failure in the Swedish deep repository is discussed. Data from studies of fission gas release from representative Swedish BWR fuel are used to estimate the average fission gas release for the spent fuel population. Information from a variety of leaching studies on LWR and CANDU fuel are then reviewed as a basis for estimating the fraction of the inventory of key radionuclides that could be released preferentially (the Instant Release Fraction of IRF) upon failure of the fuel cladding. The uncertainties associated with these estimates are discussed. 33 refs, 6 figs, 3 tabs.

  11. Macstor system for spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

    Pattantyus, P. (Atomic Energy of Canada Ltd., Montreal, PQ (Canada). Power Projects)

    1993-01-01

    In 1989, Transnuclear Inc. and AECL jointly developed the conceptual design for the Modular Aircooled Canister Storage System (Macstor) for LWR fuel. The development effort has proceeded to the completion of successful full-scale thermal testing. In 1990, AECL adapted the Macstor System approach for use with Candu fuel. The adapted design, called Canstor, has also successfully completed full-scale thermal testing, and the final system design has been completed. (author) 1 fig.

  12. Critical experiments supporting close proximity water storage of power reactor fuel. Technical progress report, October 1, 1977-December 31, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, M.N.; Hoovler, G.S.

    1978-03-01

    Experiments are being conducted on critical configurations of clusters of fuel rods, mocking up LWR-type fuel elements in close proximity water storage. Spacings between fuel clusters and the intervening material are being varied to provide a variety of benchmark loadings. (DLC)

  13. Nuclear fuels - Present and future

    Science.gov (United States)

    Olander, D.

    2009-06-01

    The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

  14. Criticality benchmark guide for light-water-reactor fuel in transportation and storage packages

    Energy Technology Data Exchange (ETDEWEB)

    Lichtenwalter, J.J.; Bowman, S.M.; DeHart, M.D.; Hopper, C.M.

    1997-03-01

    This report is designed as a guide for performing criticality benchmark calculations for light-water-reactor (LWR) fuel applications. The guide provides documentation of 180 criticality experiments with geometries, materials, and neutron interaction characteristics representative of transportation packages containing LWR fuel or uranium oxide pellets or powder. These experiments should benefit the U.S. Nuclear Regulatory Commission (NRC) staff and licensees in validation of computational methods used in LWR fuel storage and transportation concerns. The experiments are classified by key parameters such as enrichment, water/fuel volume, hydrogen-to-fissile ratio (H/X), and lattice pitch. Groups of experiments with common features such as separator plates, shielding walls, and soluble boron are also identified. In addition, a sample validation using these experiments and a statistical analysis of the results are provided. Recommendations for selecting suitable experiments and determination of calculational bias and uncertainty are presented as part of this benchmark guide.

  15. PIE of nuclear grade SiC/SiC flexural coupons irradiated to 10 dpa at LWR temperature

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    Silicon carbide fiber-reinforced SiC matrix (SiC/SiC) composites are being actively investigated for accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this study examined SiC/SiC composites following neutron irradiation at 230–340°C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials are chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC)-coated Hi-NicalonTM Type-S (HNS), TyrannoTM SA3 (SA3), and SCS-UltraTM (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young’s modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young’s moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. This study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

  16. Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Folsom, Charles Pearson [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States); Veeraraghavan, Swetha [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-05-01

    One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

  17. Evaluation of inorganic sorbent treatment for LWR coolant process streams

    Energy Technology Data Exchange (ETDEWEB)

    Roddy, J.W.

    1984-03-01

    This report presents results of a survey of the literature and of experience at selected nuclear installations to provide information on the feasibility of replacing organic ion exchangers with inorganic sorbents at light-water-cooled nuclear power plants. Radioactive contents of the various streams in boiling water reactors and pressurized water reactors were examined. In addition, the methods and performances of current methods used for controlling water quality at these plants were evaluated. The study also includes a brief review of the physical and chemical properties of selected inorganic sorbents. Some attributes of inorganic sorbents would be useful in processing light water reactor (LWR) streams. The inorganic resins are highly resistant to damage from ionizing radiation, and their exchange capacities are generally equivalent to those of organic ion exchangers. However, they are more limited in application, and there are problems with physical integrity, especially in acidic solutions. Research is also needed in the areas of selectivity and anion removal before inorganic sorbents can be considered as replacements for the synthetic organic resins presently used in LWRs. 11 figures, 14 tables.

  18. Robot Kinematics Identification: KUKA LWR4+ Redundant Manipulator Example

    Science.gov (United States)

    Kolyubin, Sergey; Paramonov, Leonid; Shiriaev, Anton

    2015-11-01

    This work is aimed at a comprehensive discussion of algorithms for the kinematic parameters identification of robotic manipulators. We deal with an open-loop geometric calibration task, when a full 6D robot's end-effector pose is measured. Effective solutions of such a task is of high interest in many practical applications, because it can dramatically improve key robot characteristics. On the first step, we select optimal calibration configurations. A comparative analysis of three different algorithms and two observability indexes used for numerical optimization is provided. Afterwards, using the acquired and pre-processed experimental data we identify modified Denavit-Hartenberg parameters of the manipulator. Estimates are obtained resolving original nonlinear forward kinematics relations. Finally, we compare nominal and calibrated geometric parameters and show how much deviations in these parameters affect robot positioning accuracy. To the best of our knowledge, such integrated efforts are new for the KUKA LWR4+ robot and Nikon K610 optical coordinate measuring machine (CMM), which were used in the study. Discussion of practical issues on how to organise the experiment is an additional contribution of this work. The proposed procedure is highly automated and can be implemented to improve manipulator's performance on a periodic basis.

  19. Democratic People`s Republic of Korea LWR project status

    Energy Technology Data Exchange (ETDEWEB)

    Mulligan, J.B.

    1996-07-01

    In October 1994, at Geneva, the United States and the Democratic People`s Republic of Korea (DPRK) signed an Agreed Framework as a first step toward resolving international concerns about nuclear activities in the DPRK. This Agreement, when implemented, will ultimately lead to the complete dismantlement of those aspects of the DPRK`s nuclear program, including reprocessing-related facilities, that have undermined the viability of the international nuclear non-proliferation regime and the stability of the Asia-Pacific region. The essence of the Agreement is that the DPRK will take near-term action to cease the activities of concern and permit some International Atomic Energy Agency (IAEA) verification inspection. In the future, it will dismantle its production reactors and accept full-scope IAWA safeguards. In return, the United Stated agreed to lead an international effort to supply the DPRK with light-water reactors which are less of proliferation concern than are graphite-moderated production reactors. Until the first LWR is in operation the DPRK will receive shipments of heavy oil to replace the energy lost by shutting down the production reactors.

  20. Impacts of transient heat transfer modeling on prediction of advanced cladding fracture during LWR LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@kaist.ac.kr; Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2016-03-15

    Highlights: • Use of constant heat transfer coefficient for fracture analysis is not sound. • On-time heat transfer coefficient should be used for thermal fracture prediction. • ∼90% of the actual fracture stresses were predicted with the on-time transient h. • Thermal-hydraulic codes can be used to better predict brittle cladding fracture. • Effects of surface oxides on thermal shock fracture should be accounted by h. - Abstract: This study presents the importance of coherency in modeling thermal-hydraulics and mechanical behavior of a solid for an advanced prediction of cladding thermal shock fracture. In water quenching, a solid experiences dynamic heat transfer rate evolutions with phase changes of the fluid over a short quenching period. Yet, such a dynamic change of heat transfer rates has been overlooked in the analysis of thermal shock fracture. In this study, we are presenting quantitative evidence against the prevailing use of a constant heat transfer coefficient for thermal shock fracture analysis in water. We conclude that no single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials. The presented results show a remarkable stress prediction improvement up to 80–90% of the actual stress with the use of the surface temperature dependent heat transfer coefficient. For thermal shock fracture analysis of brittle fuel cladding such as oxidized zirconium-based alloy or silicon carbide during LWR reflood, transient subchannel heat transfer coefficients obtained from a thermal-hydraulics code should be used as input for stress analysis. Such efforts will lead to a fundamental improvement in thermal shock fracture predictability over the current experimental empiricism for cladding fracture analysis during reflood.

  1. Assessment of reactivity transient experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ozer, O.; Yang, R.L.; Rashid, Y.R.; Montgomery, R.O.

    1996-03-01

    A few recent experiments aimed at determining the response of high-burnup LWR fuel during a reactivity initiated accident (RIA) have raised concerns that existing failure criteria may be inappropriate for such fuel. In particular, three experiments (SPERT CDC-859, NSRR HBO-1 and CABRI REP Na-1) appear to have resulted in fuel failures at only a fraction of the anticipated enthalpy levels. In evaluating the results of such RIA simulation experiments, however, it is necessary that the following two key considerations be taken into account: (1) Are the experiments representative of conditions that LWR fuel would experience during an in-reactor RIA event? (2) Is the fuel that is being utilized in the tests representative of the present (or anticipated) population of LWR fuel? Conducting experiments under conditions that can not occur in-reactor can trigger response modes that could not take place during in-reactor operation. Similarly, using unrepresentative fuel samples for the tests will produce failure information that is of limited relevance to commercial LWR fuel. This is particularly important for high-burnup fuel since the manner under which the test samples are base-irradiated prior to the test will impact the mechanical properties of the cladding and will therefore affect the RIA response. A good example of this effect can be seen in the results of the SPERT CDC-859 test and in the NSRR JM-4 and JM-5 tests. The conditions under which the fuel used for these tests was fabricated and/or base-irradiated prior to the RIA pulse resulted in the formation of multiple cladding defects in the form of hydride blisters. When this fuel was subjected to the RIA power pulse, it failed by developing multiple cracks that were closely correlated with the locations of the pre-existing hydride blisters. In the case of the JM tests, many of the cracks formed within the blisters themselves and did not propagate beyond the heavily hydrided regions.

  2. Improved LWR Cladding Performance by EPD Surface Modification Technique

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Sridharan, Kumar

    2012-11-26

    This project will utilize the electro-phoretic deposition technique (EPD) in conjunction with nanofluids to deposit oxide coatings on prototypic zirconium alloy cladding surfaces. After demonstrating that this surface modification is reproducible and robust, the team will subject the modified surface to boiling and corrosion tests to characterize the improved nucleate boiling behavior and superior corrosion performance. The scope of work consists of the following three tasks: The first task will employ the EPD surface modification technique to coat the surface of a prototypic set of zirconium alloy cladding tube materials (e.g. Zircaloy and advanced alloys such as M5) with a micron-thick layer of zirconium oxide nanoparticles. The team will characterize the modified surface for uniformity using optical microscopy and scanning-electron microscopy, and for robustness using standard hardness measurements. After zirconium alloy cladding samples have been prepared and characterized using the EPD technique, the team will begin a set of boiling experiments to measure the heat transfer coefficient and critical heat flux (CHF) limit for each prepared sample and its control sample. This work will provide a relative comparison of the heat transfer performance for each alloy and the surface modification technique employed. As the boiling heat transfer experiments begin, the team will also begin corrosion tests for these zirconium alloy samples using a water corrosion test loop that can mimic light water reactor (LWR) operational environments. They will perform extended corrosion tests on the surface-modified zirconium alloy samples and control samples to examine the robustness of the modified surface, as well as the effect on surface oxidation

  3. Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; R. Sonat Sen; Brian Boer; Abderrafi M. Ougouag; Gilles Youinou

    2011-09-01

    The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.

  4. Status report on assessment of environmentally assisted fatigue for LWR extended service conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, W. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-07-09

    This report provides an update on an earlier assessment of environmentally assisted fatigue for light water reactor (LWR) materials under extended service conditions. This report is a deliverable in September 2013, under the work package for environmentally assisted fatigue in the Light Water Reactor Sustainability (LWRS) program. The overall objective of this LWRS project is to assess the degradation by environmentally assisted cracking/fatigue of LWR materials, such as various alloy base metals and their welds used in reactor coolant system piping. This effort is to support the U.S. Department of Energy LWRS program for developing tools to predict the aging/failure mechanism and to correspondingly predict the remaining life of LWR components for anticipated 60-80 year operation.

  5. Effects of Listening While Reading (LWR on Swahili Reading Fluency and Comprehension

    Directory of Open Access Journals (Sweden)

    Filipo Lubua

    2016-10-01

    Full Text Available A number of studies have examined the contribution of technology in teaching such languages as English, French, and Spanish, among many others. Contrarily, most LCTL’s, have received very little attention. This study investigates if listening while reading (LWR may expedite Swahili reading fluency and comprehension. The study employed the iBook Author tool to create weekly mediated and interactive reading texts, with comprehension exercises, which were eventually used to collect descriptive and qualitative data from four Elementary Swahili students. Participants participated in a seven week reading program, which provided them with some kind of directed self-learning, and met with the instructor for at least 30 minutes every week for observation and more reading activities. The teacher recorded their reading scores, and a number of themes on how LWR influenced reading fluency and comprehension are discussed here. It shows that participants have a positive attitude towards LWR and they suggest it for all the reading classes.

  6. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

  7. Matpro--version 10: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    Reymann, G.A. (comp.)

    1978-02-01

    The materials properties correlations and computer subcodes (MATPRO--Version 10) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory are described. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

  8. Estimates of early containment loads from core melt accidents. Draft report for comment

    Energy Technology Data Exchange (ETDEWEB)

    None

    1985-12-01

    The thermal-hydraulic processes and corium debris-material interactions that can result from core melting in a severe accident have been studied to evaluate the potential effect of such phenomena on containment integrity. Pressure and temperature loads associated with representative accident sequences have been estimated for the six various LWR containment types used within the United States. Summaries distilling the analyses are presented and an interpretation of the results provided. 13 refs., 68 figs., 39 tabs.

  9. Radiation protection potential of MOX-fuel doped with 231Pa and Cs radioisotopes.

    Science.gov (United States)

    Kryuchkov, E F; Glebov, V B; Apse, V A; Shmelev, A N

    2005-01-01

    The paper addresses the problem of MOX-fuel self-protection during full cycle of MOX-fuel management. Under conditions of the closed LWR cycle the proliferation-resistance levels were evaluated for fresh and spent MOX-fuel with 231Pa and Cs feed. As it follows from the paper results, combination of these two admixtures being doped into MOX-fuel is able to enhance the inherent radiation barrier and to weaken shortcomings of both proliferation deterrents.

  10. Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

    Energy Technology Data Exchange (ETDEWEB)

    Miller, W.J.

    1980-05-01

    This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan.

  11. Macstor dry spent fuel storage system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. E. [Atomic Energy of Canada Limited, Montreal (Canada)

    1996-04-15

    AECL, a Canadian Grown Corporation established since 1952, is unique among the world's nuclear organizations. It is both supplier of research reactors and heavy water moderated CANDU power reactors as well as operator of extensive nuclear research facilities. As part of its mandate, AECL has developed products and conceptual designs for the short, intermediate and long term storage and disposal of spent nuclear fuel. AECL has also assumed leadership in the area of dry storage of spent fuel. This Canadian Crown Corporation first started to look into dry storage for the management of its spent nuclear fuel in the early 1970's. After developing silo-like structures called concrete canisters for the storage of its research reactor enriched uranium fuel, AECL went on to perfect that technology for spent CANDU natural uranium fuel. In 1989 AECL teamed up with Trans nuclear, Inc.,(TN), a US based member of the international Trans nuclear Group, to extend its dry storage technology to LWR spent fuel. This association combines AECL's expertise and many years experience in the design of spent fuel storage facilities with TN's proven capabilities of processing, transportation, storage and handling of LWR spent fuel. From the early AECL-designed unventilated concrete canisters to the advanced MACSTOR concept - Modular Air-Cooled Canister Storage - now available also for LWR fuel - dry storage is proving to be safe, economical, practical and, most of all, well accepted by the general public. AECL's experience with different fuels and circumstances has been conclusive.

  12. Characteristic test technology for PWR fuel and its components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho; Jeong, Yong Hwan; Park, Sang Yoon; Kim, Kyeng Ho; Nam, Cheol; Baek, Jong Hyuk; Lee, Myung Ho; Choi, Byoung Kwon; Song, Kun Woo; Kang, Ki Won; Kim, Keon Sik; Kim, Jong Hun; Kim, Young Min; Yang, Jae Ho; Song, Kee Nam; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Chun, Tae Hyun; In, Wang Kee; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-01-01

    Characteristic tests of fuel assembly and its components being developed in the Advanced LWR Fuel Development Project supported by the mid-long term nuclear R and D program are described in this report. Performance verification of fuel and its components by the characteristic tests are essential to their development. Fuel components being developed in the Advanced LWR Fuel Development Project are zirconium alloy cladding, UO{sub 2} and burnable absorber pellets, spacer grid and top and bottom end pieces. Detailed test plans for those fuel components are described in this report, and test procedures of cladding and pellet are also described in the Appendix. Examples of the described tests are in- and out-of- pile corrosion and mechanical tests such as creep and burst tests for the cladding, in-pile capsule and ramp tests for the pellet, mechanical tests such as strength and vibration, and thermal-hydraulic tests such as pressure drop and critical heat flux for the spacer grid and top and bottom end pieces. It is expected that this report could be used as the standard reference for the performance verification tests in the development of LWR fuel and its components. 11 refs., 9 figs., 2 tabs. (Author)

  13. Determination of optimal LWR containment design, excluding accidents more severe than Class 8

    Energy Technology Data Exchange (ETDEWEB)

    Cave, L.; Min, T.K.

    1980-04-01

    Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment.

  14. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

  15. Advanced Fuels Campaign Cladding & Coatings Meeting Summary

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2013-03-01

    The Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) organized a Cladding and Coatings operational meeting February 12-13, 2013, at Oak Ridge National Laboratory (ORNL). Representatives from the U.S. Department of Energy (DOE), national laboratories, industry, and universities attended the two-day meeting. The purpose of the meeting was to discuss advanced cladding and cladding coating research and development (R&D); review experimental testing capabilities for assessing accident tolerant fuels; and review industry/university plans and experience in light water reactor (LWR) cladding and coating R&D.

  16. Laser melting of uranium carbides

    Science.gov (United States)

    Utton, C. A.; De Bruycker, F.; Boboridis, K.; Jardin, R.; Noel, H.; Guéneau, C.; Manara, D.

    2009-03-01

    In the context of the material research aimed at supporting the development of nuclear plants of the fourth Generation, renewed interest has recently arisen in carbide fuels. A profound understanding of the behaviour of nuclear materials in extreme conditions is of prime importance for the analysis of the operation limits of nuclear fuels, and prediction of possible nuclear reactor accidents. In this context, the main goal of the present paper is to demonstrate the feasibility of laser induced melting experiments on stoichiometric uranium carbides; UC, UC1.5 and UC2. Measurements were performed, at temperatures around 3000 K, under a few bars of inert gas in order to minimise vaporisation and oxidation effects, which may occur at these temperatures. Moreover, a recently developed investigation method has been employed, based on in situ analysis of the sample surface reflectivity evolution during melting. Current results, 2781 K for the melting point of UC, 2665 K for the solidus and 2681 K for the liquidus of U2C3, 2754 K for the solidus and 2770 K for the liquidus of UC2, are in fair agreement with early publications where the melting behaviour of uranium carbides was investigated by traditional furnace melting methods. Further information has been obtained in the current research about the non-congruent (solidus-liquidus) melting of certain carbides, which suggest that a solidus-liquidus scheme is followed by higher ratio carbides, possibly even for UC2.

  17. Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gilles Youinou; R. Sonat Sen

    2013-09-01

    The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

  18. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ, Seoul (Korea, Republic of)

    2015-10-15

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  19. Development and experimental verification of SST-GRASS: a steady-state and transient fuel response and fission-product release code. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Rest, J.; Seitz, M.G.; Gehl, S.M.; Kelman, L.R.

    1976-01-01

    A comprehensive fission-product release model (GRASS), based on a mechanistic understanding of fuel behavior in LWR fuel elements for a wide range of accidental overheating conditions as well as steady-state irradiations, is being developed at Argonne National Laboratory. Experimental support for GRASS is provided by out-of-reactor transient heating of irradiated commercial LWR fuel using a direct-electrical-heating technique. The GRASS calculations are described, benchmarked against standard theoretical treatments, and verified for steady-state irradiations. In addition, preliminary results from the direct-electrical-heating experiments are reported. Possible mechanisms for fission-gas release during transient heating of LWR fuel are discussed based on comparisons of GRASS results with experimental observations.

  20. Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2011-11-01

    The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

  1. Neutronic evaluation of coating and cladding materials for accident tolerant fuels

    OpenAIRE

    Younker, I; Fratoni, M

    2016-01-01

    © 2015 Elsevier Ltd. All rights reserved. In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding inter...

  2. A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels

    Science.gov (United States)

    Chopra, O. K.; Rao, A. S.

    2011-02-01

    Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods changes the microstructure (radiation hardening) and microchemistry (radiation-induced segregation) of these steels, and degrades their fracture properties. Irradiation-assisted stress corrosion cracking (IASCC) is another degradation process that affects LWR internal components exposed to neutron radiation. The existing data on irradiated austenitic SSs were reviewed to evaluate the effects of key parameters such as material composition, irradiation dose, and water chemistry on IASCC susceptibility and crack growth rates of these materials in LWR environments. The significance of microstructural and microchemistry changes in the material on IASCC susceptibility is also discussed. The results are used to determine (a) the threshold fluence for IASCC and (b) the disposition curves for cyclic and IASCC growth rates for irradiated SSs in LWR environments.

  3. Development of a Reliable Fuel Depletion Methodology for the HTR-10 Spent Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kiwhan [Los Alamos National Laboratory; Beddingfield, David H. [Los Alamos National Laboratory; Geist, William H. [Los Alamos National Laboratory; Lee, Sang-Yoon [unaffiliated

    2012-07-03

    A technical working group formed in 2007 between NNSA and CAEA to develop a reliable fuel depletion method for HTR-10 based on MCNPX and to analyze the isotopic inventory and radiation source terms of the HTR-10 spent fuel. Conclusions of this presentation are: (1) Established a fuel depletion methodology and demonstrated its safeguards application; (2) Proliferation resistant at high discharge burnup ({approx}80 GWD/MtHM) - Unfavorable isotopics, high number of pebbles needed, harder to reprocess pebbles; (3) SF should remain under safeguards comparable to that of LWR; and (4) Diversion scenarios not considered, but can be performed.

  4. Safety and Regulatory Issues of the Thorium Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian [ORNL; Worrall, Andrew [ORNL; Powers, Jeffrey [ORNL; Bowman, Steve [ORNL; Flanagan, George [ORNL; Gehin, Jess [ORNL

    2014-02-01

    Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2), add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.

  5. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  6. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  7. A Stochastic LWR Model with Consideration of the Driver's Individual Property%A Stochastic LWR Model with Consideration of the Driver's Individual Property

    Institute of Scientific and Technical Information of China (English)

    唐铁桥; 王云鹏; 余贵珍; 黄海军

    2012-01-01

    In this paper,we develop a stochastic LWR model based on the influences of the driver's individual property on his/her perceived density and speed deviation.The numerical results show that the driver's individual property has great effects on traffic flow only when the initial density is moderate,i.e.,at this time,oscillating traffic flow will occur and the oscillating phenomena in the traffic system consisting of the conservative and aggressive drivers is more serious than that in the traffic system consisting of the conservative (aggressive) drivers.

  8. Assessment of LWR spent fuel disposal options. Volume 2. System definition and assessments. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-01

    Volume 2, in addition to introducing and defining the requirements for the study, contains the detailed evaluation of the study cases in each of six assessment areas, and describes and evaluates a number of study variations.

  9. High Temperature and Pressure Steam-H2 Interaction with Candidate Advanced LWR Fuel Claddings

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A [ORNL

    2012-08-01

    This report summarizes the work completed to evaluate cladding materials that could serve as improvements to Zircaloy in terms of accident tolerance. This testing involved oxidation resistance to steam or H{sub 2}-50% steam environments at 800-1350 C at 1-20 bar for short times. A selection of conventional alloys, SiC-based ceramics and model alloys were used to explore a wide range of materials options and provide guidance for future materials development work. Typically, the SiC-based ceramic materials, alumina-forming alloys and Fe-Cr alloys with {ge}25% Cr showed the best potential for oxidation resistance at {ge}1200 C. At 1350 C, FeCrAl alloys and SiC remained oxidation resistant in steam. Conventional austenitic steels do not have sufficient oxidation resistance with only {approx}18Cr-10Ni. Higher alloyed type 310 stainless steel is protective but Ni is not a desirable alloy addition for this application and high Cr contents raise concern about {alpha}{prime} formation. Higher pressures (up to 20.7 bar) and H{sub 2} additions appeared to have a limited effect on the oxidation behavior of the most oxidation resistant alloys but higher pressures accelerated the maximum metal loss for less oxidation resistant steels and less metal loss was observed in a H{sub 2}-50%H{sub 2}O environment at 10.3 bar. As some of the results regarding low-alloyed FeCrAl and Fe-Cr alloys were unexpected, further work is needed to fundamentally understand the minimum Cr and Al alloy contents needed for protective behavior in these environments in order to assist in alloy selection and guide alloy development.

  10. Evaluation of Methods for Decladding LWR Fuel for a Pyroprocessing-Based Reprocessing Plant

    Science.gov (United States)

    1992-10-01

    removed prior to any other operation that has the potential to contaminate the plenum structure. The compression springs, typically made of Inconel - 718 ...44 A.1.4 Dissolution in Hydrofluoric Acid .............................. 46 A.1.5 Nonaqueous Dissolution ...46 A.1.6 Alcholic HCl Solutions ..................................... 48 A.1.7 Direct Dissolution in Molten Lead or Zinc Chloride

  11. Precharacterization Report for Instrumented Fuel Assembly (IFA)-527

    Energy Technology Data Exchange (ETDEWEB)

    Cunningham, M. E.; Bradley, E. R.; Daniel, J. L.; Davis, N. C.; Lanning, D. D.; Williford, R. E.

    1981-07-01

    This report is a resource document covering the rationale, design, fabrication, and preirradiation characterization of instrumented fuel assembly (IFA)-527. This assembly is being irradiated in the Halden Boiling Water Reactor (HBWR) in Norway as part of the Experimental Support and Development of Single-Rod Fuel Codes Program conducted by Pacific Northwest laboratory (PNL) and sponsored by the Fuel Behavior Research Branch of the U.S. Nuclear Regulatory Commission (NRC). Data from this assembly will be used to better understand light water reactor (LWR) fuel behavior under normal operating conditions.

  12. Validation of the BISON 3D Fuel Performance Code: Temperature Comparisons for Concentrically and Eccentrically Located Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Hales; D. M. Perez; R. L. Williamson; S. R. Novascone; B. W. Spencer

    2013-03-01

    BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. The code is applicable to both steady and transient fuel behaviour and is used to analyse either 2D axisymmetric or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods. Halden IFA experiments constitute a large percentage of the current BISON validation base. The validation emphasis here is centreline temperatures at the beginning of fuel life, with comparisons made to seven rods from the IFA-431 and 432 assemblies. The principal focus is IFA-431 Rod 4, which included concentric and eccentrically located fuel pellets. This experiment provides an opportunity to explore 3D thermomechanical behaviour and assess the 3D simulation capabilities of BISON. Analysis results agree with experimental results showing lower fuel centreline temperatures for eccentric fuel with the peak temperature shifted from the centreline. The comparison confirms with modern 3D analysis tools that the measured temperature difference between concentric and eccentric pellets is not an artefact and provides a quantitative explanation for the difference.

  13. Draft report: a selection methodology for LWR safety R and D programs and proposals

    Energy Technology Data Exchange (ETDEWEB)

    Husseiny, A. A.; Ritzman, R. L.

    1980-03-01

    The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application.

  14. JAEA key facilities for global advanced fuel cycle R and D

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Shigeo; Yamamoto, Ryuichi [Nuclear Fuel Cycle Engineering Labos, JAEA, 4-33 Tokai-mura, Ibaraki, 319-1194 (Japan)

    2008-07-01

    Advanced fuel cycle will be realized with the mid and long term R and D during the long-term transition period from LWR cycle to advanced reactor fuel cycle. Most of JAEA facilities have been utilized to establish the current LWR and FBR (Fast Breeder Reactor) fuel cycle by implementing evolutionary R and D. An assessment of today's state experimental facilities concerning the following research issues: reprocessing, Mox fuel fabrication, irradiation and post-irradiation examination, waste management and nuclear data measurement, is made. The revolutionary R and D requests new issues to be studied: the TRU multi-recycling, minor actinide recycling, the assessment of proliferation resistance and the assessment of cost reduction. To implement the revolutionary R and D for advanced fuel cycle, however, these facilities should be refurbished to install new machines and process equipment to provide more flexible testing parameters.

  15. Results of the LIRES Round Robin test on high temperature reference electrodes for LWR applications

    Energy Technology Data Exchange (ETDEWEB)

    Bosch, R.W. [SCK.CEN, Nuclear Research Centre Belgium, Boeretang 200, B-2400 Mol (Belgium); Nagy, G. [Magyar Tudomanyos Akademia KFKI Atomenergia Kutatointezet, AEKI, Konkoly Thege ut 29-33, 1121 Budapest (Hungary); Feron, D. [CEA Saclay, 91191 Gif-Sur-Yvette Cedex (France); Navas, M. [CIEMAT, Edificio 30, Dpto. Fision Nuclear, Avda. Complutense 22, 28040 Madrid, (Spain); Bogaerts, W. [KU Leuven, Kasteelpark Arenberg 31, B-3001 Leuven (Belgium); Karnik, D. [Nuclear Research Institute, NRI, Rez (Czech Republic); Dorsch, T. [Framatone ANP, Inc., Charlotte, North Carolina (United States); Molander, A. [Studsvik AB SE-611 82 Nykoeping (Sweden); Maekelae, K. [Materials and Structural Integrity, VTT Technical Research Centre of Finland, Kemistintie 3, P.O. Box 1704, FIN-02044 VTT (Finland)

    2004-07-01

    A European sponsored research project has been started on 1 October 2000 to develop high temperature reference electrodes that can be used for in-core electrochemical measurements in Light Water Reactors (LWR's). This LIRES-project (Development of Light Water Reactor Reference Electrodes) consists of 9 partners (SCK-CEN, AEKI, CEA, CIEMAT, KU Leuven, NRI Rez, Framatone ANP, Studsvik Nuclear and VTT) and will last for four years. The main objective of this LIRES project is to develop a reference electrode, which is robust enough to be used inside a LWR. Emphasize is put on the radiation hardness of both the mechanical design of the electrode as the proper functioning of the electrode. A four steps development trajectory is foreseen: (1) To set a testing standard for a Round Robin, (2) To develop different reference electrodes, (3) To perform a Round Robin test of these reference electrodes followed by selection of the best reference electrode(s), (4) To perform irradiation tests under appropriate LWR conditions in a Material Test Reactor (MTR). Four different high temperature reference electrodes have been developed and are being tested in a Round Robin test. These electrodes are: A Ceramic Membrane Electrode (CME), a Rhodium electrode, an external Ag/AgCl electrode and a Palladium electrode. The presentation will focus on the results obtained with the Round Robin test. (authors)

  16. Analysis of advanced european nuclear fuel cycle scenarios including transmutation and economical estimates

    Energy Technology Data Exchange (ETDEWEB)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F. [CIEMAT, Avda. Complutense, 40, 28040 Madrid (Spain)

    2013-07-01

    In this work the transition from the existing Light Water Reactors (LWR) to the advanced reactors is analyzed, including Generation III+ reactors in a European framework. Four European fuel cycle scenarios involving transmutation options have been addressed. The first scenario (i.e., reference) is the current fleet using LWR technology and open fuel cycle. The second scenario assumes a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel. The third scenario is a modification of the second one introducing Minor Actinide (MA) transmutation in a fraction of the FR fleet. Finally, in the fourth scenario, the LWR fleet is replaced using FR with MOX fuel as well as Accelerator Driven Systems (ADS) for MA transmutation. All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for a period of 200 years looking for equilibrium mass flows. The simulations were made using the TR-EVOL code, a tool for fuel cycle studies developed by CIEMAT. The results reveal that all scenarios are feasible according to nuclear resources demand (U and Pu). Concerning to no transmutation cases, the second scenario reduces considerably the Pu inventory in repositories compared to the reference scenario, although the MA inventory increases. The transmutation scenarios show that elimination of the LWR MA legacy requires on one hand a maximum of 33% fraction (i.e., a peak value of 26 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation). On the other hand a maximum number of ADS plants accounting for 5% of electricity generation are predicted in the fourth scenario (i.e., 35 ADS units). Regarding the economic analysis, the estimations show an increase of LCOE (Levelized cost of electricity) - averaged over the whole period - with respect to the reference scenario of 21% and 29% for FR and FR with transmutation scenarios respectively, and 34% for the fourth scenario. (authors)

  17. Chemical state of tellurium in a degraded LWR core

    Science.gov (United States)

    Imoto, S.; Tanabe, T.

    1988-06-01

    Changes of the chemical state of tellurium in the heatup stage of a severe fuel damage accident are estimated thermodynamically. According to equilibrium calculations with the SOLGASMIX-PV code, tellurium exists as cesium telluride, as the element or possibly as PdTe during normal operation. In the heatup stage of an accident, elemental tellurium is absorbed in the Zircaloy cladding by formation of ZrTe x ( x = 1-2). Cesium telluride does not react with Zr even under the low oxygen potentials favoring the {Zr}/{UO 2} reaction. Tellurium is also absorbed in oxygen-stabilized alpha-zirconium. The stability of Cs 2Te in the steam/hydrogen atmosphere is discussed.

  18. Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

    Science.gov (United States)

    Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

    2005-02-01

    A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

  19. A comparison of mechanical algorithms of fuel performance code systems

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

    2003-11-01

    The goal of fuel rod performance evaluation is to identify the robustness of fuel rod with cladding material during fuel irradiation. Computer simulation of fuel rod performance becomes important to develop new nuclear systems. To construct the computing code system for fuel rod performance, we compared several algorithms of existing fuel rod performance code systems and summarized the details and tips as a preliminary work. Among several code systems, FRAPCON, FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. The computational algorithms related to mechanical interaction of the fuel rod are compared including methodologies and subroutines. This work will be utilized to develop the computing code system for dry process fuel rod performance.

  20. Experiments and analyses on melt-structure-water interactions during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Seghal, B.R.; Dinh, T.N.; Bui, V.A.; Green, J.A.; Nourgaliev, R.R.; Okkonen, T.O.; Dinh, A.T. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-04-01

    This report is the final report for the research project Melt Structure Water Interactions (MSWI). It describes results of analytical and experimental studies concerning MSWI during the course of a hypothetical core meltdown accident in a LWR. Emphasis has been placed on phenomena which govern vessel failure mode and timing and the mechanisms and properties which govern the fragmentation and breakup of melt jets and droplets. It was found that: 2-D effects significantly diminished the focusing effect of an overlying metallic layer on top of an oxide melt pool. This result improves the feasibility of in-vessel retention of a melt pool through external cooling of the lower head; phenomena related to hole ablation and melt discharge, in the event of vessel failure, are affected significantly by crust formation; the jet fragmentation process is a function of many related phenomena. The fragmentation rate depends not only on the traditional parameters but also on the melt physical properties, which change as the melt cools down from liquid to solid temperature; film boiling was investigated by developing a two-phase flow model and inserting it in a multi-D fluid dynamics code. It was concluded that the thickness of the film on the surface of a melt jet would be small and that the effects of the film on the process should not be large. This conclusion is contrary to the modeling employed in some other codes. The computer codes were developed and validated against the data obtained in the MSWI Project. The melt vessel interaction thermal analysis code describes the process of melt pool formation and convection and the resulting vessel thermal loadings. In addition, several innovative models were developed to describe the melt-water interaction process. The code MELT-3D treats the melt jet as a collection of particles whose movement is described with a three-dimensional Eulerian formulation. The model (SIPHRA) tracks the melt jet with an additional equation, using the

  1. MCOR - Monte Carlo depletion code for reference LWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)

    2011-04-15

    Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally

  2. Technology of the light water reactor fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Wymer, R. G.

    1979-01-01

    This essay presents elements of the processes used in the fuel cycle steps and gives an indication of the types of equipment used. The amounts of radioactivity released in normal operation of the processes are indicated and related to radiation doses. Types and costs of equipment or processes required to lower these radioactivity releases are in some cases suggested. Mining and milling, conversion of uranium concentrate to UF/sub 6/, uranium isotope separation, LWR fuel fabrication, fuel reprocessing, transportation, and waste management are covered in this essay. 40 figures, 34 tables. (DLC)

  3. Cold spray deposition of Ti2AlC coatings for improved nuclear fuel cladding

    Science.gov (United States)

    Maier, Benjamin R.; Garcia-Diaz, Brenda L.; Hauch, Benjamin; Olson, Luke C.; Sindelar, Robert L.; Sridharan, Kumar

    2015-11-01

    Coatings of Ti2AlC MAX phase compound have been successfully deposited on Zircaloy-4 (Zry-4) test flats, with the goal of enhancing the accident tolerance of LWR fuel cladding. Low temperature powder spray process, also known as cold spray, has been used to deposit coatings ∼90 μm in thickness using powder particles of accident tolerance to nuclear fuel cladding.

  4. RADIOACTIVE WASTE STREAMS FROM VARIOUS POTENTIAL NUCLEAR FUEL CYCLE OPTIONS

    Energy Technology Data Exchange (ETDEWEB)

    Nick Soelberg; Steve Piet

    2010-11-01

    Five fuel cycle options, about which little is known compared to more commonly known options, have been studied in the past year for the United States Department of Energy. These fuel cycle options, and their features relative to uranium-fueled light water reactor (LWR)-based fuel cycles, include: • Advanced once-through reactor concepts (Advanced Once-Through, or AOT) – intended for high uranium utilization and long reactor operating life, use depleted uranium in some cases, and avoid or minimize used fuel reprocessing • Fission-fusion hybrid (FFH) reactor concepts – potential variations are intended for high uranium or thorium utilization, produce fissile material for use in power generating reactors, or transmute transuranic (TRU) and some radioactive fission product (FP) isotopes • High temperature gas reactor (HTGR) concepts - intended for high uranium utilization, high reactor thermal efficiencies; they have unique fuel designs • Molten salt reactor (MSR) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, use on-line reprocessing of the used fuel, produce lesser amounts of long-lived, highly radiotoxic TRU elements, and avoid fuel assembly fabrication • Thorium/U-233 fueled LWR (Th/U-233) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, and produce lesser amounts of long-lived, highly radiotoxic TRU elements. These fuel cycle options could result in widely different types and amounts of used or spent fuels, spent reactor core materials, and waste streams from used fuel reprocessing, such as: • Highly radioactive, high-burnup used metal, oxide, or inert matrix U and/or Th fuels, clad in Zr, steel, or composite non-metal cladding or coatings • Spent radioactive-contaminated graphite, SiC, carbon-carbon-composite, metal, and Be reactor core materials • Li-Be-F salts containing U, TRU, Th, and fission products • Ranges of separated or un-separated activation

  5. Validation of the THIRMAL-1 melt-water interaction code

    Energy Technology Data Exchange (ETDEWEB)

    Chu, C.C.; Sienicki, J.J.; Spencer, B.W. [Argonne National Lab., IL (United States)

    1995-09-01

    The THIRMAL-1 computer code has been used to calculate nonexplosive LWR melt-water interactions both in-vessel and ex-vessel. To support the application of the code and enhance its acceptability, THIRMAL-1 has been compared with available data from two of the ongoing FARO experiments at Ispra and two of the Corium Coolant Mixing (CCM) experiments performed at Argonne. THIRMAL-1 calculations for the FARO Scoping Test and Quenching Test 2 as well as the CCM-5 and -6 experiments were found to be in excellent agreement with the experiment results. This lends confidence to the modeling that has been incorporated in the code describing melt stream breakup due to the growth of both Kelvin-Helmholtz and large wave instabilities, the sizes of droplets formed, multiphase flow and heat transfer in the mixing zone surrounding and below the melt metallic phase. As part of the analysis of the FARO tests, a mechanistic model was developed to calculate the prefragmentation as it may have occurred when melt relocated from the release vessel to the water surface and the model was compared with the relevant data from FARO.

  6. Technical program to study the benefits of nonlinear analysis methods in LWR component designs. Technical report TR-3723-1

    Energy Technology Data Exchange (ETDEWEB)

    Raju, P. P.

    1980-05-01

    This report summarizes the results of the study program to assess the benefits of nonlinear analysis methods in Light Water Reactor (LWR) component designs. The current study reveals that despite its increased cost and other complexities, nonlinear analysis is a practical and valuable tool for the design of LWR components, especially under ASME Level D service conditions (faulted conditions) and it will greatly assist in the evaluation of ductile fracture potential of pressure boundary components. Since the nonlinear behavior is generally a local phenomenon, the design of complex components can be accomplished through substructuring isolated localized regions and evaluating them in detail using nonlinear analysis methods.

  7. Transient fuel behavior of preirradiated PWR fuels under reactivity initiated accident conditions

    Science.gov (United States)

    Fujishiro, Toshio; Yanagisawa, Kazuaki; Ishijima, Kiyomi; Shiba, Koreyuki

    1992-06-01

    Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

  8. LWR decay heat calculations using a GRS improved ENDF/B-6 based ORIGEN data library

    Energy Technology Data Exchange (ETDEWEB)

    Hesse, U.; Hummelsheim, K.I.; Kilger, R.; Moser, F.E.; Langenbuch, S. [Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, Garching (Germany)

    2008-07-01

    The known ORNL ORIGEN code is widely spread over the world for inventory, activity and decay heat tasks and is used stand-alone or implemented in activation, shielding or burn-up systems. More than 1000 isotopes with more than six coupled neutron capture and radioactive decay channels are handled simultaneously by the code. The characteristics of the calculated inventories, e.g., masses, activities, neutron and photon source terms or the decay heat during short or long decay time steps are achieved by summing over all isotopes, characterized in the ORIGEN libraries. An extended nuclear GRS-ORIGENX data library is now developed for practical appliance. The library was checked for activation tasks of structure material isotopes and for actinide and fission product burn-up calculations compared with experiments and standard methods. The paper is directed to the LWR decay heat calculation features of the new library and shows the differences of dynamical and time integrated results of Endf/B-6 based and older Endf/B-5 based libraries for decay heat tasks compared to fission burst experiments, ANS curves and some other published data. A multi-group time exponential evaluation is given for the fission burst power of {sup 235}U, {sup 238}U, {sup 239}Pu and {sup 241}Pu, to be used in quick LWR reactor accident decay heat calculation tools. (authors)

  9. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Shack, W.J. [Argonne National Lab., IL (United States)

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented.

  10. On the Diffusion Coefficient of Two-step Method for LWR analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Deokjung; Choi, Sooyoung [UNIST, Ulsan (Korea, Republic of); Smith, Kord S. [Massachusetts Institute of Technology, Cambridge (United States)

    2015-10-15

    The few-group constants including diffusion coefficients are generated from the assembly calculation results. Once the assembly calculation is done, the cross sections (XSs) are spatially homogenized, and a critical spectrum calculation is performed in order to take into account the neutron leakages of the lattice. The diffusion coefficient is also generated through the critical spectrum calculation. Three different methods of the critical spectrum calculation such as B1 method, P1 method, and fundamental mode (FM) calculation method are considered in this paper. The diffusion coefficients can also be affected by transport approximations for the transport XS calculation which is used in the assembly transport lattice calculation in order to account for the anisotropic scattering effects. The outflow transport approximation and the inflow transport approximation are investigated in this paper. The accuracy of the few group data especially the diffusion coefficients has been studied to optimize the combination of the transport correction methods and the critical spectrum calculation methods using the UNIST lattice physics code STREAM. The combination of the inflow transport approximation and the FM method is shown to provide the highest accuracy in the LWR core calculations. The methodologies to calculate the diffusion coefficients have been reviewed, and the performances of them have been investigated with a LWR core problem. The combination of the inflow transport approximation and the fundamental mode critical spectrum calculation shows the smallest errors in terms of assembly power distribution.

  11. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    Energy Technology Data Exchange (ETDEWEB)

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water

  12. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors

    Science.gov (United States)

    Yamamoto, Y.; Pint, B. A.; Terrani, K. A.; Field, K. G.; Yang, Y.; Snead, L. L.

    2015-12-01

    Development of nuclear grade, iron-based wrought FeCrAl alloys has been initiated for light water reactor (LWR) fuel cladding to serve as a substitute for zirconium-based alloys with enhanced accident tolerance. Ferritic alloys with sufficient chromium and aluminum additions can exhibit significantly improved oxidation kinetics in high-temperature steam environments when compared to zirconium-based alloys. In the first phase, a set of model FeCrAl alloys containing 10-20Cr, 3-5Al, and 0-0.12Y in weight percent, were prepared by conventional arc-melting and hot-working processes to explore the effect of composition on the properties of FeCrAlY alloys. It was found that the tensile properties were insensitive to the alloy compositions studied; however, the steam oxidation resistance strongly depended on both the chromium and the aluminum contents. The second phase development focused on strengthening Fe-13Cr-5Al with minor alloying additions of molybdenum, niobium, and silicon. Combined with an optimized thermo-mechanical treatment, a thermally stable microstructure was produced with improved tensile properties at temperatures up to 741 °C.

  13. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yanghyun; Kim, Keonsik; Park, Jeongyong; Yang, Yongsik; Kim, Hyungkyu; In, Wangkee; Song, Kunwoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-03-15

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR.

  14. Fuel flexible fuel injector

    Science.gov (United States)

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  15. Spent fuel characteristics & disposal considerations

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M.

    1996-06-01

    The fuel used in commercial nuclear power reactors is uranium, generally in the form of an oxide. The gas-cooled reactors developed in England use metallic uranium enclosed in a thin layer of Magnox. Since this fuel must be processed into a more stable form before disposal, we will not consider the characteristics of the Magnox spent fuel. The vast majority of the remaining power reactors in the world use uranium dioxide pellets in Zircaloy cladding as the fuel material. Reactors that are fueled with uranium dioxide generally use water as the moderator. If ordinary water is used, the reactors are called Light Water Reactors (LWR), while if water enriched in the deuterium isotope of hydrogen is used, the reactors are called Heavy Water reactors. The LWRs can be either pressurized reactors (PWR) or boiling water reactors (BWR). Both of these reactor types use uranium that has been enriched in the 235 isotope to about 3.5 to 4% total abundance. There may be minor differences in the details of the spent fuel characteristics for PWRs and BWRs, but for simplicity we will not consider these second-order effects. The Canadian designed reactor (CANDU) that is moderated by heavy water uses natural uranium without enrichment of the 235 isotope as the fuel. These reactors run at higher linear power density than LWRs and produce spent fuel with lower total burn-up than LWRs. Where these difference are important with respect to spent fuel management, we will discuss them. Otherwise, we will concentrate on spent fuel from LWRs.

  16. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  17. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

    2017-10-02

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  18. Nonlinear analysis of LWR components: areas of investigation/benefits/recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, S. J. [ed.

    1980-04-01

    The purpose of this study is to identify specific topics of investigation into design procedures, design concepts, methods of analysis, testing practices, and standards which are characterized by nonlinear behavior (both geometric and material) and which are considered to offer some economic and/or technical benefits to the LWR industry (excluding piping). In this study these topics were collected, compiled, and subjectively evaluated as to their potential benefit. The topics considered to have the greatest benefit/impact potential are discussed. The topics of investigation were found to fall basically into three areas: component, code interpretation, and load/failure mechanism. The topics are arbitrarily reorganized into six areas of investigation: Fracture, Fatigue, Vibration/Dynamic/Seismic, Plasticity, Component/Computational Considerations, and Code Interpretation.

  19. Melting of Transition Metals

    Energy Technology Data Exchange (ETDEWEB)

    Ross, M; Japel, S; Boehler, R

    2005-04-11

    We review the transition melting studies carried out at Mainz, and describe a recently developed model used to explain that the relatively low melting slopes are due to the partially filled d-bands, and the persistence of the pressure induced s-d transition. The basic tenets of the model have now been reconfirmed by new measurements for Cu and Ni. The measurements show that Cu which has a filled 3d-band, has a melt slope that is about 2.5 greater than its neighbor Ni. In the case of Mo, the apparent discrepancy of DAC melting measurements with shock melting can be explained by accounting for the change in melt slope due to the bcc-cp transition observed in the shock studies. The Fe melt curve is revisited. The possible relevance of the Jahn-Teller effect and recently observed transition metal melts with Icosahedral Short-Range Order (ISRO) is discussed.

  20. 轻水堆乏燃料和钍燃料在ACR-700利用的探索%Investigation of Potential Utilization for Light Water Reactor Spent Fuel and Thorium Fuel in ACR-700

    Institute of Scientific and Technical Information of China (English)

    邹春燕; 陈金根; 蔡翔舟; 蒋大真; 郭锐; 陈堃; 郭威; 马余刚; 胡碧涛

    2013-01-01

    轻水堆乏燃料和钍燃料的利用是解决乏燃料后处理问题和核燃料短缺的有效途径之一.本工作以ACR 700标准燃料为参考,研究了4种不同混合比例的轻水堆乏燃料及钍燃料在ACR-700中的k∞和燃耗.研究结果表明,将裂变产物分离后,轻水堆乏燃料的重锕系核素在ACR-700中可作为一很好的燃料;只要加入足够的启动燃料,钍燃料也可作为很好的转换燃料,使反应堆内生成233U的速率大于易裂变燃料的消耗速率,233U的生成对反应堆运行后期维持临界起重要作用.%The potential utilization of light water reactor (LWR) spent fuel and thorium fuel is a very efficient way for solving spent fuel processing and nuclear fuel shortage problems.Four different mixed fuel types with LWR spent fuel and Th fuel were researched in ACR-700 using SCALE5.1 code system.Compared with the ACR-700 standard fuel,the fuel cycle model 1 and 4 can reach higher burnup.According to the research on the lattice physics of ACR-700 fuel bundle,it is indicated that LWR spent fuel after separation of the fission products is a very prospective fuel in ACR-700 and 232Th can be a preferable fertile isotope if sufficient booster fuel is supplied.

  1. Performance of Trasuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Interim Report, Including Void Reactivity Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; Brian Boer; Gilles Youinou; Abderrafi M. Ougouag

    2011-03-01

    The current focus of the Deep Burn Project is on once-through burning of transuranice (TRU) in light water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles would be pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell calculations have been performed using the DRAGON-4 code in order assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells containing typical UO2 and MOX fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Loading of TRU-only FCM fuel into a pin without significant quantities of uranium challenges the design from the standpoint of several key reactivity parameters, particularly void reactivity, and to some degree, the Doppler coefficient. These unit cells, while providing an indication of how a whole core of similar fuel would behave, also provide information of how individual pins of TRU-only FCM fuel would influence the reactivity behavior of a heterogeneous assembly. If these FCM fuel pins are included in a heterogeneous assembly with LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance of the TRU-only FCM fuel pins may be preserved. A configuration such as this would be similar to CONFU assemblies analyzed in previous studies. Analogous to the plutonium content limits imposed on MOX fuel, some amount of TRU-only FCM pins in an otherwise-uranium fuel assembly may give acceptable reactivity

  2. Modeling and Analysis of FCM UN TRISO Fuel Using the PARFUME Code

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2013-09-01

    The PARFUME (PARticle Fuel ModEl) modeling code was used to assess the overall fuel performance of uranium nitride (UN) tri-structural isotropic (TRISO) ceramic fuel in the frame of the design and development of Fully Ceramic Matrix (FCM) fuel. A specific modeling of a TRISO particle with UN kernel was developed with PARFUME, and its behavior was assessed in irradiation conditions typical of a Light Water Reactor (LWR). The calculations were used to access the dimensional changes of the fuel particle layers and kernel, including the formation of an internal gap. The survivability of the UN TRISO particle was estimated depending on the strain behavior of the constituent materials at high fast fluence and burn-up. For nominal cases, internal gas pressure and representative thermal profiles across the kernel and layers were determined along with stress levels in the pyrolytic carbon (PyC) and silicon carbide (SiC) layers. These parameters were then used to evaluate fuel particle failure probabilities. Results of the study show that the survivability of UN TRISO fuel under LWR irradiation conditions might only be guaranteed if the kernel and PyC swelling rates are limited at high fast fluence and burn-up. These material properties are unknown at the irradiation levels expected to be reached by UN TRISO fuel in LWRs. Therefore, more effort is needed to determine them and positively conclude on the applicability of FCM fuel to LWRs.

  3. Conceptual design and economic analysis of a light water reactor fuel enricher/regenerator. FY 1978 year-end report

    Energy Technology Data Exchange (ETDEWEB)

    Grand, P; Kouts, H J; Powell, J R; Steinberg, M; Takahashi, H

    1979-05-01

    A study has been performed to evaluate the use of high-energy particle accelerators as nuclear fuel enrichers and nuclear fuel regenerators. This builds on ideas that have been current for many years. The new study has, however, explored some novel approaches that have not been examined before. A specific conceptual system chosen for more detailed study would stretch the energy available from natural uranium by a factor of about 3, reduce the separative work requirements by a factor of about 4, and reduce the volume of spent fuel to be stored by a factor of 2, compared to the current once-through light water reactor (LWR) fuel cycle. The concept avoids the need for chemical reprocessing of spent fuel, and would permit continued use of LWR's beyond the time when limitations on fuel resources might otherwise lead to their being phased out. This concept, which is called the Linear Accelerator Fuel Enricher/Regenerator, is therefore viewed as offering a practical means of stretching the use of the nuclear fuel resource in the framework of the existing light water reactor fuel cycle. This report describes and analyzes the concept referred to. An explanation of the principles underlying the concept is given. Particular attention is devoted to engineering feasibility, proliferation resistance, and economics. It is seen that the concept draws on only proven technology as regards bothaccelerator design and the fuel irradiation process, and is adapted to existing LWR designs with no change except in fuel-handling practices. A preliminary evaluation of radiation damage, coolant options, and power conversion systems is provided. Neutronic, thermal-hydraulic, and burnup calculations are presented. An analysis is made of fuel economy. Approximate costs of electric power produced using this concept are evaluated and discussed. Estimated development costs of commercialization are provided.

  4. Sustainability Features of Nuclear Fuel Cycle Options

    Directory of Open Access Journals (Sweden)

    Stefano Passerini

    2012-09-01

    Full Text Available The nuclear fuel cycle is the series of stages that nuclear fuel materials go through in a cradle to grave framework. The Once Through Cycle (OTC is the current fuel cycle implemented in the United States; in which an appropriate form of the fuel is irradiated through a nuclear reactor only once before it is disposed of as waste. The discharged fuel contains materials that can be suitable for use as fuel. Thus, different types of fuel recycling technologies may be introduced in order to more fully utilize the energy potential of the fuel, or reduce the environmental impacts and proliferation concerns about the discarded fuel materials. Nuclear fuel cycle systems analysis is applied in this paper to attain a better understanding of the strengths and weaknesses of fuel cycle alternatives. Through the use of the nuclear fuel cycle analysis code CAFCA (Code for Advanced Fuel Cycle Analysis, the impact of a number of recycling technologies and the associated fuel cycle options is explored in the context of the U.S. energy scenario over 100 years. Particular focus is given to the quantification of Uranium utilization, the amount of Transuranic Material (TRU generated and the economics of the different options compared to the base-line case, the OTC option. It is concluded that LWRs and the OTC are likely to dominate the nuclear energy supply system for the period considered due to limitations on availability of TRU to initiate recycling technologies. While the introduction of U-235 initiated fast reactors can accelerate their penetration of the nuclear energy system, their higher capital cost may lead to continued preference for the LWR-OTC cycle.

  5. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  6. Status of the MELTSPREAD-1 computer code for the analysis of transient spreading of core debris melts

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.; Chu, C.C.

    1992-01-01

    A transient, one dimensional, finite difference computer code (MELTSPREAD-1) has been developed to predict spreading behavior of high temperature melts flowing over concrete and/or steel surfaces submerged in water, or without the effects of water if the surface is initially dry. This paper provides a summary overview of models and correlations currently implemented in the code, code validation activities completed thus far, LWR spreading-related safety issues for which the code has been applied, and the status of documentation for the code.

  7. Status of the MELTSPREAD-1 computer code for the analysis of transient spreading of core debris melts

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.; Chu, C.C.

    1992-04-01

    A transient, one dimensional, finite difference computer code (MELTSPREAD-1) has been developed to predict spreading behavior of high temperature melts flowing over concrete and/or steel surfaces submerged in water, or without the effects of water if the surface is initially dry. This paper provides a summary overview of models and correlations currently implemented in the code, code validation activities completed thus far, LWR spreading-related safety issues for which the code has been applied, and the status of documentation for the code.

  8. Evaluation of CASMO-3 and HELIOS for Fuel Assembly Analysis from Monte Carlo Code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hyung Jin; Song, Jae Seung; Lee, Chung Chan

    2007-05-15

    This report presents a study comparing deterministic lattice physics calculations with Monte Carlo calculations for LWR fuel pin and assembly problems. The study has focused on comparing results from the lattice physics code CASMO-3 and HELIOS against those from the continuous-energy Monte Carlo code McCARD. The comparisons include k{sub inf}, isotopic number densities, and pin power distributions. The CASMO-3 and HELIOS calculations for the k{sub inf}'s of the LWR fuel pin problems show good agreement with McCARD within 956pcm and 658pcm, respectively. For the assembly problems with Gadolinia burnable poison rods, the largest difference between the k{sub inf}'s is 1463pcm with CASMO-3 and 1141pcm with HELIOS. RMS errors for the pin power distributions of CASMO-3 and HELIOS are within 1.3% and 1.5%, respectively.

  9. Nuclear Material Attractiveness: An Assessment of Material from PHWR's in a Closed Thorium Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Sleaford, Brad W.; Ebbinghaus, B. B.; Bradley, Keith S.; Robel, Martin; Prichard, Andrew W.; Smith, Brian W.; Collins, Brian A.; Hase, Kevin R.; Jarvinen, G. D.; Ireland, J. R.; Johnson, M. W.; Bathke, Charles G.; Wallace, R. K.

    2010-06-11

    This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies [ , ] that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that 233U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [ ]. The methodology and key findings will be presented.

  10. Ice Thermal Storage Systems for LWR Supplemental Cooling and Peak Power Shifting

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Hongbin Zhang; Phil Sharpe; Blaise Hamanaka; Wei Yan; WoonSeong Jeong

    2010-06-01

    Availability of enough cooling water has been one of the major issues for the nuclear power plant site selection. Cooling water issues have frequently disrupted the normal operation at some nuclear power plants during heat waves and long draught. The issues become more severe due to the new round of nuclear power expansion and global warming. During hot summer days, cooling water leaving a power plant may become too hot to threaten aquatic life so that environmental regulations may force the plant to reduce power output or even temporarily to be shutdown. For new nuclear power plants to be built at areas without enough cooling water, dry cooling can be used to remove waste heat directly into the atmosphere. However, dry cooling will result in much lower thermal efficiency when the weather is hot. One potential solution for the above mentioned issues is to use ice thermal storage systems (ITS) that reduce cooling water requirements and boost the plant’s thermal efficiency in hot hours. ITS uses cheap off-peak electricity to make ice and uses those ice for supplemental cooling during peak demand time. ITS is suitable for supplemental cooling storage due to its very high energy storage density. ITS also provides a way to shift large amount of electricity from off peak time to peak time. Some gas turbine plants already use ITS to increase thermal efficiency during peak hours in summer. ITSs have also been widely used for building cooling to save energy cost. Among three cooling methods for LWR applications: once-through, wet cooling tower, and dry cooling tower, once-through cooling plants near a large water body like an ocean or a large lake and wet cooling plants can maintain the designed turbine backpressure (or condensation temperature) during 99% of the time; therefore, adding ITS to those plants will not generate large benefits. For once-through cooling plants near a limited water body like a river or a small lake, adding ITS can bring significant economic

  11. Development of metallic fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Ho; Lee, Chong Yak; Lee, Myung Ho and others

    1999-03-01

    With the vacuum melting and casting of the U-10wt%Zr alloy which is metallic fuel for liquid metal fast breeder reactor, we studied the microstructure of the alloy and the parameters of the melting and casting for the fuel rods. Internal defects of the U-10wt%Zr fuel by gravity casting, were inspected by non-destructive test. U-10wt%Zr alloy has been prepared for the thermal stability test in order to estimate the decomposition of the lamellar structure with relation to swelling under irradiation condition. (author)

  12. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

    Directory of Open Access Journals (Sweden)

    Lindley Benjamin A.

    2016-01-01

    Full Text Available The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO2/PuO2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs-clad systems, particularly for current and near-term build LWRs. R&D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN and uranium silicide (U3Si2. Candidate cladding materials include advanced stainless steel (FeCrAl and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R&D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I2S-LWR, a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP Integrated Research Project (IRP is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I2S-LWR design are U3Si2 and advanced stainless steel respectively. In addition, the I2S-LWR design

  13. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  14. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  15. Uniform corrosion of FeCrAl alloys in LWR coolant environments

    Science.gov (United States)

    Terrani, K. A.; Pint, B. A.; Kim, Y.-J.; Unocic, K. A.; Yang, Y.; Silva, C. M.; Meyer, H. M.; Rebak, R. B.

    2016-10-01

    The corrosion behavior of commercial and model FeCrAl alloys and type 310 stainless steel was examined by autoclave tests and compared to Zircaloy-4, the reference cladding materials in light water reactors. The corrosion studies were carried out in three distinct water chemistry environments found in pressurized and boiling water reactor primary coolant loop conditions for up to one year. The structure and morphology of the oxides formed on the surface of these alloys was consistent with thermodynamic predictions. Spinel-type oxides were found to be present after hydrogen water chemistry exposures, while the oxygenated water tests resulted in the formation of very thin and protective hematite-type oxides. Unlike the alloys exposed to oxygenated water tests, the alloys tested in hydrogen water chemistry conditions experienced mass loss as a function of time. This mass loss was the result of net sum of mass gain due to parabolic oxidation and mass loss due to dissolution that also exhibits parabolic kinetics. The maximum thickness loss after one year of LWR water corrosion in the absence of irradiation was ∼2 μm, which is inconsequential for a ∼300-500 μm thick cladding.

  16. Proceedings of the IAEA specialists` meeting on cracking in LWR RPV head penetrations

    Energy Technology Data Exchange (ETDEWEB)

    Pugh, C.E.; Raney, S.J. [comps.] [Oak Ridge National Lab., TN (United States)

    1996-07-01

    This report contains 17 papers that were presented in four sessions at the IAEA Specialists` meeting on Cracking in LWR RPV Head Penetrations held at ASTM Headquarters in Philadelphia on May 2-3, 1995. The papers are compiled here in the order that presentations were made in the sessions, and they relate to operational observations, inspection techniques, analytical modeling, and regulatory control. The goal of the meeting was to allow international experts to review experience in the field of ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. The emphasis was to allow a better understanding of RPV material behavior, to provide guidance supporting reliability and adequate performance, and to assist in defining directions for further investigations. The international nature of the meeting is illustrated by the fact that papers were presented by researchers from 10 countries. There were technical experts present form other countries who participated in discussions of the results presented. This present document incorporates the final version of the papers as received from the authors. The final chapter includes conclusions and recommendations. Individual papers have been cataloged separately.

  17. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S. [and others

    1995-09-01

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

  18. Implementation and evaluation of fuel creep using advanced light-water reactor materials in FRAPCON 3.5

    Science.gov (United States)

    Carroll, Spencer

    As current reactors approach the end of their operable lifetime, new reactors are needed if nuclear power is to continue being generated in the United States. Some utilities have already began construction on newer, more advanced LWR reactors, which use the same fuel as current reactors and have a similar but updated design. Others are researching next generation (GEN-IV) reactors which have new designs that utilize alternative fuel, coolants and other reactor materials. Many of these alternative fuels are capable of achieving higher burnups and are designed to be more accident tolerant than the currently used UO2 fuel. However, before these new materials can be used, extensive research must be done in order to obtain a detailed understanding of how the new fuels and other materials will interact. New fuels, such as uranium nitride (UN) and uranium carbide (UC) have several advantages over UO2, such as increased burnup capabilities and higher thermal conductivities. However, there are issues with each that prevent UC and UN from being used as direct replacements for UO2. Both UC and UN swell at a significantly higher rate than UO2 and neither fuel reacts favorably when exposed to water. Due to this, UC and UN are being considered more for GEN-IV reactors that use alternative coolant rather than for current LWRs. In an effort to increase accident tolerance, silicon carbide (SiC) is being considered for use as an alternative cladding. The high strength, high melting point and low oxidation of SiC make it an attractive cladding choice, especially in an accident scenario. However, as a ceramic, SiC is not ductile and will not creep outwards upon pellet-clad mechanical interaction (PCMI) which can cause a large build up in interfacial pressure. In order to understand the interaction between the high swelling fuels and unyielding SiC cladding, data on the properties and behaviors of these materials must be gathered and incorporated into FRAPCON. FRAPCON is a fuel

  19. Assessment of LWR spent fuel disposal options. Volume 3. Study bases and system design considerations (Appendices). Technical report

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-01

    Volume 3 (Appendices) provides a tabulation of the bases and assumptions used in the study as well as preconceptual design description and cost estimates of the facilities and transportation systems necessary to implement the various study cases.

  20. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 5. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Volume V of the five-volume report consists of appendices, which provide supplementary information, with emphasis on characteristics of geologic formations that might be used for final storage or disposal. Appendix titles are: selected glossary; conversion factors; geologic isolation, including, (a) site selection factors for repositories of wastes in geologic media, (b) rock types--geologic occurrence, (c) glossary of geohydrologic terms, and (d) 217 references; the ocean floor; and, government regulations pertaining to the management of radioactive materials. (JGB)

  1. Evaluation Metrics Applied to Accident Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton; Jon Carmack; Frank Goldner

    2014-10-01

    The safe, reliable, and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and have yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. One of the current missions of the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) is to develop nuclear fuels and claddings with enhanced accident tolerance for use in the current fleet of commercial LWRs or in reactor concepts with design certifications (GEN-III+). Accident tolerance became a focus within advanced LWR research upon direction from Congress following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of ATF development is to identify alternative fuel system technologies to further enhance the safety, competitiveness and economics of commercial nuclear power. Enhanced accident tolerant fuels would endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system while maintaining or improving performance during normal operations. The U.S. DOE is supporting multiple teams to investigate a number of technologies that may improve fuel system response and behavior in accident conditions, with team leadership provided by DOE national laboratories, universities, and the nuclear industry. Concepts under consideration offer both evolutionary and revolutionary changes to the current nuclear fuel system. Mature concepts will be tested in the Advanced Test Reactor at Idaho National Laboratory beginning in Summer 2014 with additional concepts being

  2. Fuel Cycle System Analysis Handbook

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Brent W. Dixon; Dirk Gombert; Edward A. Hoffman; Gretchen E. Matthern; Kent A. Williams

    2009-06-01

    This Handbook aims to improve understanding and communication regarding nuclear fuel cycle options. It is intended to assist DOE, Campaign Managers, and other presenters prepare presentations and reports. When looking for information, check here. The Handbook generally includes few details of how calculations were performed, which can be found by consulting references provided to the reader. The Handbook emphasizes results in the form of graphics and diagrams, with only enough text to explain the graphic, to ensure that the messages associated with the graphic is clear, and to explain key assumptions and methods that cause the graphed results. Some of the material is new and is not found in previous reports, for example: (1) Section 3 has system-level mass flow diagrams for 0-tier (once-through), 1-tier (UOX to CR=0.50 fast reactor), and 2-tier (UOX to MOX-Pu to CR=0.50 fast reactor) scenarios - at both static and dynamic equilibrium. (2) To help inform fast reactor transuranic (TRU) conversion ratio and uranium supply behavior, section 5 provides the sustainable fast reactor growth rate as a function of TRU conversion ratio. (3) To help clarify the difference in recycling Pu, NpPu, NpPuAm, and all-TRU, section 5 provides mass fraction, gamma, and neutron emission for those four cases for MOX, heterogeneous LWR IMF (assemblies mixing IMF and UOX pins), and a CR=0.50 fast reactor. There are data for the first 10 LWR recycle passes and equilibrium. (4) Section 6 provides information on the cycle length, planned and unplanned outages, and TRU enrichment as a function of fast reactor TRU conversion ratio, as well as the dilution of TRU feedstock by uranium in making fast reactor fuel. (The recovered uranium is considered to be more pure than recovered TRU.) The latter parameter impacts the required TRU impurity limits specified by the Fuels Campaign. (5) Section 7 provides flows for an 800-tonne UOX separation plant. (6) To complement 'tornado' economic

  3. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    Energy Technology Data Exchange (ETDEWEB)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W. [and others

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program.

  4. Signatures of nonthermal melting

    Directory of Open Access Journals (Sweden)

    Tobias Zier

    2015-09-01

    Full Text Available Intense ultrashort laser pulses can melt crystals in less than a picosecond but, in spite of over thirty years of active research, for many materials it is not known to what extent thermal and nonthermal microscopic processes cause this ultrafast phenomenon. Here, we perform ab-initio molecular-dynamics simulations of silicon on a laser-excited potential-energy surface, exclusively revealing nonthermal signatures of laser-induced melting. From our simulated atomic trajectories, we compute the decay of five structure factors and the time-dependent structure function. We demonstrate how these quantities provide criteria to distinguish predominantly nonthermal from thermal melting.

  5. Application of Thermochemical Modeling to Assessment/Evaluation of Nuclear Fuel Behavior

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, Theodore M [University of South Carolina, Columbia; McMurray, Jake W [ORNL; Simunovic, Srdjan [ORNL

    2016-01-01

    The combination of new fuel compositions and higher burn-ups envisioned for the future means that representing fuel properties will be much more important, and yet more complex. Behavior within the oxide fuel rods will be difficult to model owing to the high temperatures, and the large number of elements generated and their significant concentrations that are a result of fuels taken to high burn-up. This unprecedented complexity offers an enormous challenge to the thermochemical understanding of these systems and opportunities to advance solid solution models to describe these materials. This paper attempts to model and simulate that behavior using an oxide fuels thermochemical description to compute the equilibrium phase state and oxygen potential of LWR fuel under irradiation.

  6. Melt Cast High Explosives

    Directory of Open Access Journals (Sweden)

    Stanisław Cudziło

    2014-12-01

    Full Text Available [b]Abstract[/b]. This paper reviews the current state and future developments of melt-cast high explosives. First the compositions, properties and methods of preparation of trinitrotoluene based (TNT conventional mixtures with aluminum, hexogen (RDX or octogen (HMX are described. In the newer, less sensitive explosive formulations, TNT is replaced with dinitroanisole (DNANDNANDNAN and nitrotriazolone (NTONTONTO, nitroguanidine (NG or ammonium perchlorate (AP are the replacement for RDRDX and HMX. Plasticized wax or polymer-based binder systems for melt castable explosives are also included. Hydroxyl terminated polybutadiene (HPTB is the binder of choice, but polyethylene glycol, and polycaprolactone with energetic plasticizers are also used. The most advanced melt-cast explosives are compositions containing energetic thermoplastic elastomers and novel highly energetic compounds (including nitrogen rich molecules in whose particles are nanosized and practically defect-less.[b]Keywords[/b]: melt-cast explosives, detonation parameters

  7. Melting of sodium clusters

    CERN Document Server

    Reyes-Nava, J A; Beltran, M R; Michaelian, K

    2002-01-01

    Thermal stability properties and the melting-like transition of Na_n, n=13-147, clusters are studied through microcanonical molecular dynamics simulations. The metallic bonding in the sodium clusters is mimicked by a many-body Gupta potential based on the second moment approximation of a tight-binding Hamiltonian. The characteristics of the solid-to-liquid transition in the sodium clusters are analyzed by calculating physical quantities like caloric curves, heat capacities, and root-mean-square bond length fluctuations using simulation times of several nanoseconds. Distinct melting mechanisms are obtained for the sodium clusters in the size range investigated. The calculated melting temperatures show an irregular variation with the cluster size, in qualitative agreement with recent experimental results. However, the calculated melting point for the Na_55 cluster is about 40 % lower than the experimental value.

  8. Advanced Fuels Campaign FY 2014 Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    Braase, Lori [Idaho National Lab. (INL), Idaho Falls, ID (United States). INL Systems Analyses; May, W. Edgar [Idaho National Lab. (INL), Idaho Falls, ID (United States). INL Systems Analyses

    2014-10-01

    The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address the longer-term waste management challenges. This includes development of a state-of-the art Research and Development (R&D) infrastructure to support the use of a “goal-oriented science-based approach.” In support of the Fuel Cycle Research and Development (FCRD) program, AFC is responsible for developing advanced fuels technologies to support the various fuel cycle options defined in the Department of Energy (DOE) Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. AFC uses a “goal-oriented, science-based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. The modeling and simulation activities for fuel performance are carried out under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which is closely coordinated with AFC. In this report, the word “fuel” is used generically to include fuels, targets, and their associated cladding materials. R&D of light water reactor (LWR) fuels with enhanced accident tolerance is also conducted by AFC. These fuel systems are designed to achieve significantly higher fuel and plant performance to allow operation to significantly higher burnup, and to provide enhanced safety during design basis and beyond design basis accident conditions. The overarching goal is to develop advanced nuclear fuels and materials that are robust, have high performance capability, and are more tolerant to

  9. Early implementation of SiC cladding fuel performance models in BISON

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-18

    SiC-based ceramic matrix composites (CMCs) [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation due to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.

  10. Spent fuel workshop'2002

    Energy Technology Data Exchange (ETDEWEB)

    Poinssot, Ch

    2002-07-01

    This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO{sub 2} fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO{sub 2} dissolution determined from electrochemical experiments with {sup 238}Pu doped UO{sub 2} M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO{sub 2} studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with {alpha} doped UO{sub 2} in Boom clay conditions (K. Lemmens), Studies of the behavior of UO{sub 2} / water interfaces under He{sup 2+} beam (C. Corbel), Alpha and gamma radiolysis effects on UO{sub 2} alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines

  11. Force induced DNA melting

    Energy Technology Data Exchange (ETDEWEB)

    Santosh, Mogurampelly; Maiti, Prabal K [Center for Condensed Matter Theory, Department of Physics, Indian Institute of Science, Bangalore-12 (India)], E-mail: santosh@physics.iisc.ernet.in, E-mail: maiti@physics.iisc.ernet.in

    2009-01-21

    When pulled along the axis, double-strand DNA undergoes a large conformational change and elongates by roughly twice its initial contour length at a pulling force of about 70 pN. The transition to this highly overstretched form of DNA is very cooperative. Applying a force perpendicular to the DNA axis (unzipping), double-strand DNA can also be separated into two single-stranded DNA, this being a fundamental process in DNA replication. We study the DNA overstretching and unzipping transition using fully atomistic molecular dynamics (MD) simulations and argue that the conformational changes of double-strand DNA associated with either of the above mentioned processes can be viewed as force induced DNA melting. As the force at one end of the DNA is increased the DNA starts melting abruptly/smoothly above a critical force depending on the pulling direction. The critical force f{sub m}, at which DNA melts completely decreases as the temperature of the system is increased. The melting force in the case of unzipping is smaller compared to the melting force when the DNA is pulled along the helical axis. In the case of melting through unzipping, the double-strand separation has jumps which correspond to the different energy minima arising due to sequence of different base pairs. The fraction of Watson-Crick base pair hydrogen bond breaking as a function of force does not show smooth and continuous behavior and consists of plateaus followed by sharp jumps.

  12. Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance

    Directory of Open Access Journals (Sweden)

    Cheng Bo

    2016-01-01

    Full Text Available Molybdenum based alloy is selected as a candidate to enhance tolerance of fuel to severe loss of coolant accidents due to its high melting temperature of ∼2600 °C and ability to maintain sufficient mechanical strength at temperatures exceeding 1200 °C. An outer layer of either a Zr-alloy or Al-containing stainless steel is designed to provide corrosion resistance under normal operation and oxidation resistance in steam exceeding 1000 °C for 24 hours under severe loss of coolant accidents. Due to its higher neutron absorption cross-sections, the Mo-alloy cladding is designed to be less than half the thickness of the current Zr-alloy cladding. A feasibility study has been undertaken to demonstrate (1 fabricability of long, thin wall Mo-alloy tubes, (2 formability of a protective outer coating, (3 weldability of Mo tube to endcaps, (4 corrosion resistance in autoclaves with simulated LWR coolant, (5 oxidation resistance to steam at 1000–1500 °C, and (6 sufficient axial and diametral strength and ductility. High purity Mo as well as Mo + La2O3 ODS alloy have been successfully fabricated into ∼2-meter long tubes for the feasibility study. Preliminary results are encouraging, and hence rodlets with Mo-alloy cladding containing fuel pellets have been under preparation for irradiation at the Advanced Test Reactor (ATR in Idaho National Laboratory. Additional efforts are underway to enhance the Mo cladding mechanical properties via process optimization. Oxidation tests to temperatures up to 1500 °C, and burst and creep tests up to 1000 °C are also underway. In addition, some Mo disks in close contact with UO2 from a previous irradiation program (to >100 GWd/MTU at the Halden Reactor have been subjected to post-irradiation examination to evaluate the chemical compatibility of Mo with irradiated UO2 and fission products. This paper will provide an update on results from the feasibility study and discuss the attributes of the

  13. Discrete Modeling of Early-Life Thermal Fracture in Ceramic Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Huang, Hai [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dolbow, John E. [Duke Univ., Durham, NC (United States); Hales, Jason D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    Fracturing of ceramic fuel pellets heavily influences performance of light water reactor (LWR) fuel. Early in the life of fuel, starting with the initial power ramp, large thermal gradients cause high tensile hoop and axial stresses in the outer region of the fuel pellets, resulting in the formation of radial and axial cracks. Circumferential cracks form due to thermal gradients that occur when the power is ramped down. These thermal cracks cause the fuel to expand radially, closing the pellet/cladding gap and enhancing the thermal conductance across that gap, while decreasing the effective conductivity of the fuel in directions normal to the cracking. At lower length scales, formation of microcracks is an important contributor to the decrease in bulk thermal conductivity that occurs over the life of the fuel as the burnup increases. Because of the important effects that fracture has on fuel performance, a realistic, physically based fracture modeling capability is essential to predict fuel behavior in a wide variety of normal and abnormal conditions. Modeling fracture within the context of the finite element method, which is based on continuous interpolations of solution variables, has always been challenging because fracture is an inherently discontinuous phenomenon. Work is underway at Idaho National Laboratory to apply two modeling techniques model fracture as a discrete displacement discontinuity to nuclear fuel: The extended finite element method (XFEM), and discrete element method (DEM). XFEM is based on the standard finite element method, but with enhancements to represent discontinuous behavior. DEM represents a solid as a network of particles connected by bonds, which can arbitrarily fail if a fracture criterion is reached. This paper presents initial results applying the aforementioned techniques to model fuel fracturing. This work has initially focused on early life behavior of ceramic LWR fuel. A coupled thermal-mechanical XFEM method that includes

  14. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

    2015-06-01

    FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

  15. Uncertainty Analysis of Light Water Reactor Fuel Lattices

    Directory of Open Access Journals (Sweden)

    C. Arenas

    2013-01-01

    Full Text Available The study explored the calculation of uncertainty based on available cross-section covariance data and computational tool on fuel lattice levels, which included pin cell and the fuel assembly models. Uncertainty variations due to temperatures changes and different fuel compositions are the main focus of this analysis. Selected assemblies and unit pin cells were analyzed according to the OECD LWR UAM benchmark specifications. Criticality and uncertainty analysis were performed using TSUNAMI-2D sequence in SCALE 6.1. It was found that uncertainties increase with increasing temperature, while kinf decreases. This increase in the uncertainty is due to the increase in sensitivity of the largest contributing reaction of uncertainty, namely, the neutron capture reaction 238U(n, γ due to the Doppler broadening. In addition, three types (UOX, MOX, and UOX-Gd2O3 of fuel material compositions were analyzed. A remarkable increase in uncertainty in kinf was observed for the case of MOX fuel. The increase in uncertainty of kinf in MOX fuel was nearly twice the corresponding value in UOX fuel. The neutron-nuclide reaction of 238U, mainly inelastic scattering (n, n′, contributed the most to the uncertainties in the MOX fuel, shifting the neutron spectrum to higher energy compared to the UOX fuel.

  16. Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt A [ORNL; Kiggans Jr, James O [ORNL; McMurray, Jake W [ORNL; Jolly, Brian C [ORNL; Hunt, Rodney Dale [ORNL; Trammell, Michael P [ORNL; Snead, Lance Lewis [ORNL

    2016-01-01

    Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhance heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.

  17. MELTED BUTTER TECHNOLOGY DEVELOPMENT

    Directory of Open Access Journals (Sweden)

    L. V. Golubeva

    2014-01-01

    Full Text Available Summary. Melted butter is made from dairy butter by rendering the fat phase. It has specific taste and aroma, high-calorie content and good assimilability. Defects of butter which appeared during the storage causes by the development of microbiological processes or by the chemical oxidation. On the development of these processes influence quality and composition of fresh butter, its physical structure, content of the increased amount of gas phase and content of heavy metals, storage conditions. Microbiological spoilage of butter occurs generally due to damage of plasma which is good environment for the development of microorganisms. Defects of microbiological origin include: unclean, sour, moldy, yeasty, cheesy, bitter taste. Defects of test and smell chemical origin are formed due to hydrolytic digestion of lipids. It's prevailed at long storage of butter in the conditions of freezing temperatures. It's picked out the following main processes of spoiling: souring, acidifying and sallowness. Often these processes take place simultaneously.It has been investigated melted butter with lactated additive. The latter improves the microbiological and toxicological safety, prolongs the storage condition of the products. Technological efficiency of the additives is achieved by a multilayer products formation from the inactive bound water, preventing microorganisms growth and by the barrier layer with lactate inhibiting hydrolytic reactions. Oil samples were obtained with the batch-type butter maker application, then they were melted and after that lactated additive were supplemented. It has been studied organoleptic and physico-chemical indices of the melted butter samples. The fatty-acid composition of melted butter were studied. Comparative analysis of fatty-acid composition of cow's milk fat and produced melted butter has shown their similarity. Also in the last sample there is increased weight fraction of linoleic and linolenic acids. The obtained

  18. Analysis of Pu-Only Partitioning Strategies in LMFBR Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Samuel Bays; Gilles Youinou

    2013-02-01

    Sodium cooled Fast Reactors (SFR) have been under consideration for production of electricity, fissile material production, and for destruction of transuranics for decades. The neutron economy of a SFR can be operated in one of two ways. One possibility is to operate the reactor in a transuranic burner mode which has been the focus of active R&D in the last 15 years. However, prior to that the focus was on breeding transuranics. This later mode of managing the neutron economy relies on ensuring the maximum fuel utilization possible in such a way as to maximize the amount of plutonium produced per unit of fission energy in the reactor core. The goal of maximizing plutonium production in this study is as fissile feed stock for the production of MOX fuel to be used in Light Water Reactors (LWR). Throughout the l970’s, this fuel cycle scenario was the focus of much research by the Atomic Energy Commission in the event that uranium supplies would be scarce. To date, there has been sufficient uranium to supply the once through nuclear fuel cycle. However, interest in a synergistic relationship Liquid Metal Fast Breeder Reactors (LMFBR) and a consumer LWR fleet persists, prompting this study. This study considered LMFBR concepts with varying additions of axial and radial reflectors. Three scenarios were considered in collaboration with a companion study on the LWR-MOX designs based on plutonium nuclide vectors produced by this study. The first scenario is a LMFBR providing fissile material to make MOX fuel where the MOX part of the fuel cycle is operated in a once-through-then-out mode. The second scenario is the same as the first but with the MOX part of the fuel cycle multi-recycling its own plutonium with LMFBR being used for the make-up feed. In these first two scenarios, plutonium partitioning from the minor actinides (MA) was assumed. Also, the plutonium management strategy of the LMFBR ensured that only the high fissile purity plutonium bred from blankets was

  19. Identification of the impacts of maintenance and testing upon the safety of LWR power plants. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Husseiny, A. A.; Sabri, Z. A.; Turnage, J. J.

    1980-04-01

    The present study was designed to identify the impact of maintenance and testing (M and T) upon the safety of LWR power plants. The study involved data extraction from various sources reporting safety-related and operation-related nuclear power plant experience. Primary sources reviewed, including Licensee Event Reports (LER's) submitted to the NRC, revealed that only ten percent of events reported could be identified as M and T problems. The collected data were collated in a manner that would allow identification of principal types of problems which are associated with the performance of M and T tasks in LWR power plants. Frequencies of occurrence of events and their general endemic nature were analyzed using data clustering and pattern recognition techniques, as well as chi-square analyses for sparse contingency tables. The results of these analyses identified seven major categories of M and T error modes which were related to individual facilities and reactor type. Data review indicated that few M and T problems were directly related to procedural inadequacies, with the majority of events being attributable to human error.

  20. Full Core 3-D Simulation of a Partial MOX LWR Core

    Energy Technology Data Exchange (ETDEWEB)

    S. Bays; W. Skerjanc; M. Pope

    2009-05-01

    A comparative analysis and comparison of results obtained between 2-D lattice calculations and 3-D full core nodal calculations, in the frame of MOX fuel design, was conducted. This study revealed a set of advantages and disadvantages, with respect to each method, which can be used to guide the level of accuracy desired for future fuel and fuel cycle calculations. For the purpose of isotopic generation for fuel cycle analyses, the approach of using a 2-D lattice code (i.e., fuel assembly in infinite lattice) gave reasonable predictions of uranium and plutonium isotope concentrations at the predicted 3-D core simulation batch average discharge burnup. However, it was found that the 2-D lattice calculation can under-predict the power of pins located along a shared edge between MOX and UO2 by as much as 20%. In this analysis, this error did not occur in the peak pin. However, this was a coincidence and does not rule out the possibility that the peak pin could occur in a lattice position with high calculation uncertainty in future un-optimized studies. Another important consideration in realistic fuel design is the prediction of the peak axial burnup and neutron fluence. The use of 3-D core simulation gave peak burnup conditions, at the pellet level, to be approximately 1.4 times greater than what can be predicted using back-of-the-envelope assumptions of average specific power and irradiation time.

  1. THE MISSION AND ACCOMPLISHMENTS FROM DOE’S FUEL CYCLE RESEARCH AND DEVELOPMENT (FCRD) ADVANCED FUELS CAMPAIGN

    Energy Technology Data Exchange (ETDEWEB)

    J. Carmack; L. Braase; F. Goldner

    2015-09-01

    The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors, enhance proliferation resistance of nuclear fuel, effectively utilize nuclear energy resources, and address the longer-term waste management challenges. This includes development of a state of the art Research and Development (R&D) infrastructure to support the use of a “goal oriented science based approach.” AFC uses a “goal oriented, science based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. One of the most challenging aspects of AFC is the management, integration, and coordination of major R&D activities across multiple organizations. AFC interfaces and collaborates with Fuel Cycle Technologies (FCT) campaigns, universities, industry, various DOE programs and laboratories, federal agencies (e.g., Nuclear Regulatory Commission [NRC]), and international organizations. Key challenges are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Challenged with the research and development of fuels for two different reactor technology platforms, AFC targeted transmutation fuel development and focused ceramic fuel development for Advanced LWR Fuels.

  2. The underwater coincidence counter (UWCC) for plutonium measurements in mixed oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Eccleston, G.W.; Menlove, H.O.; Abhold, M.; Baker, M.; Pecos, J.

    1998-12-31

    The use of fresh uranium-plutonium mixed oxide (MOX) fuel in light-water reactors (LWR) is increasing in Europe and Japan and it is necessary to verify the plutonium content in the fuel for international safeguards purposes. The UWCC is a new instrument that has been designed to operate underwater and nondestructively measure the plutonium in unirradiated MOX fuel assemblies. The UWCC can be quickly configured to measure either boiling-water reactor (BWR) or pressurized-water reactor (PWR) fuel assemblies. The plutonium loading per unit length is measured using the UWCC to precisions of less than 1% in a measurement time of 2 to 3 minutes. Initial calibrations of the UWCC were completed on measurements of MOX fuel in Mol, Belgium. The MCNP-REN Monte Carlo simulation code is being benchmarked to the calibration measurements to allow accurate simulations for extended calibrations of the UWCC.

  3. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  4. Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Ferrandis, J.Y.; Leveque, G.; Rosenkrantz, E.; Augereau, F.; Combette, P. [University Montpellier, IES, UMR 5214, F-34000, Montpellier (France); CNRS, IES, UMR 5214, F-34000, Montpellier (France)

    2015-07-01

    irradiation. The instrumented fuel rod incorporating the ultrasonic gas composition sensor was finally irradiated during 2 weeks in nominal conditions. Neutronics calculation will be performed in order to calculate the thermal and fast neutron fluence and the gamma dose absorbed by acoustic sensor. A first evaluation gives a thermal fluence about 4,5.10{sup 19} n/cm{sup 2}, a fast neutrons fluence about 4,5.1018 n/cm{sup 2} and a total gamma dose up to 0,25 MGy The maximal temperature during the irradiation test was about 150 C. Although the ultrasonic sensor appears to be damaged, the optimization of the electrical attack parameters and the development of a new signal processing maintain the measurement feasibility up the end of the irradiation campaign. It was the first time that the composition of fission gas has been monitored all along an irradiation experiment in a MTR, giving access to the gas release kinetics. New researches involve thick film transducers produced by screen-printing process in order to propose piezoelectric structures for harsh temperature and irradiation measurements. The second project consists in the development of a non-destructive device that can be directly applied on a LWR fuel rod. The problem to be solved relates to the measurement of the fission gas pressure and composition in a fuel rod using a non-destructive method. Fuel rod internal pressure is one of the safety criteria applied in nuclear power analyses. This criterion must be verified in order to avoid any fuel-cladding gap reopening risk and therefore any local clad ballooning. Apart from the safety implications, this parameter is also a fuel behaviour indicator and reflects the overall fuel performance in operation, but also during shipping and long-term storage. Rod internal pressure is one criterion amongst others, like cladding corrosion, against which the acceptable fuel burn-up limit is set. A sensor has been achieved in 2007. A full-scale hot cell test of the internal gas

  5. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  6. Development of fuel performance and thermal hydraulic technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

    2000-03-01

    Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

  7. GLASS MELTING PHENOMENA, THEIR ORDERING AND MELTING SPACE UTILISATION

    Directory of Open Access Journals (Sweden)

    Němec L.

    2013-12-01

    Full Text Available Four aspects of effective glass melting have been defined – namely the fast kinetics of partial melting phenomena, a consideration of the melting phenomena ordering, high utilisation of the melting space, and effective utilisation of the supplied energy. The relations were defined for the specific melting performance and specific energy consumption of the glass melting process which involve the four mentioned aspects of the process and indicate the potentials of effective melting. The quantity “space utilisation” has been treated in more detail as an aspect not considered in practice till this time. The space utilisation was quantitatively defined and its values have been determined for the industrial melting facility by mathematical modelling. The definitions of the specific melting performance and specific energy consumption have been used for assessment of the potential impact of a controlled melt flow and high space utilisation on the melting process efficiency on the industrial scale. The results have shown that even the partial control of the melt flow, leading to the partial increase of the space utilisation, may considerably increase the melting performance, whereas a decrease of the specific energy consumption was determined to be between 10 - 15 %.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Viscosity Measurement for Tellurium Melt

    Science.gov (United States)

    Lin, Bochuan; Li, Chao; Ban, Heng; Scripa, Rosalia N.; Su, Ching-Hua; Lehoczky, Sandor L.

    2006-01-01

    The viscosity of high temperature Te melt was measured using a new technique in which a rotating magnetic field was applied to the melt sealed in a suspended ampoule, and the torque exerted by rotating melt flow on the ampoule wall was measured. Governing equations for the coupled melt flow and ampoule torsional oscillation were solved, and the viscosity was extracted from the experimental data by numerical fitting. The computational result showed good agreement with experimental data. The melt velocity transient initiated by the rotating magnetic field reached a stable condition quickly, allowing the viscosity and electrical conductivity of the melt to be determined in a short period.

  10. Thermoacoustic Streaming and Ultrasonic Processing of Low Melting Melts

    Science.gov (United States)

    Trinh, E. H.

    1997-01-01

    Ultrasonic levitation allows the processing of low melting materials both in 1 G as well as in microgravity. The free suspension of the melts also facilitates undercooling, permitting the measurements of the physical properties of the metastable liquids.

  11. ASSESSMENT OF POSSIBLE CYCLE LENGTHS FOR FULLY-CERAMIC MICRO-ENCAPSULATED FUEL-BASED LIGHT WATER REACTOR CONCEPTS

    Energy Technology Data Exchange (ETDEWEB)

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal Pasamehmetoglu; Francesco Venneri

    2012-04-01

    The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o.

  12. Research on low emission MSW gasification and melting system

    Institute of Scientific and Technical Information of China (English)

    XIAO Gang; NI Mingjiang; CHI Yong; JIN Yuqi; ZHANG Jiaquan; MIAO Qi; CEN Kefa

    2007-01-01

    In order to eliminate secondary pollution caused by municipal solid waste (MSW) incineration,a MSW gasification and melting process is proposed.The process is expected to reduce the emission of pollutants,especially heavy-metals and dioxins.In this paper,the combustible components of MSW and simulated MSW were gasified in a lab-scale fluidized bed at 400℃-700℃ when the excess air ratio (ER) was between 0.2 and 0.8.The experimental results indicated that the MSW could be gasified effectively in a fluidized bed at approximately 600℃-700℃ when excess air ratio was 0.2-0.4.The melting characteristics of two typical fly ash samples from MSW incinerators were investigated.The results indicated that fly ash of pure MSW incineration could be melted at approximately 1,300℃ and that of MSW and coal co-combustion could be melted at approximately 1,400℃.When temperature was over 1,100℃,more than 99.9% of the dioxins could be decomposed and most of the heavy-metals could be solidified in the slag.Based on the above experiments,two feasible MSW gasification and melting processes were proposed for low calorific value MSW:(1) sieved MSW gasification and melting system,which was based on an idea of multi-recycle;(2) gasification and melting scheme of MSW adding coal as assistant fuel.

  13. Thermodynamics of Oligonucleotide Duplex Melting

    Science.gov (United States)

    Schreiber-Gosche, Sherrie; Edwards, Robert A.

    2009-01-01

    Melting temperatures of oligonucleotides are useful for a number of molecular biology applications, such as the polymerase chain reaction (PCR). Although melting temperatures are often calculated with simplistic empirical equations, application of thermodynamics provides more accurate melting temperatures and an opportunity for students to apply…

  14. Melting of polydisperse hard disks

    NARCIS (Netherlands)

    Pronk, S.; Frenkel, D.

    2004-01-01

    The melting of a polydisperse hard-disk system is investigated by Monte Carlo simulations in the semigrand canonical ensemble. This is done in the context of possible continuous melting by a dislocation-unbinding mechanism, as an extension of the two-dimensional hard-disk melting problem. We find

  15. Melting of polydisperse hard disks

    NARCIS (Netherlands)

    Pronk, S.; Frenkel, D.

    2004-01-01

    The melting of a polydisperse hard-disk system is investigated by Monte Carlo simulations in the semigrand canonical ensemble. This is done in the context of possible continuous melting by a dislocation-unbinding mechanism, as an extension of the two-dimensional hard-disk melting problem. We find th

  16. Development Status of Accident Tolerant Fuel Cladding for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Hydrogen explosions and the release of radionuclides are caused by severe damage of current nuclear fuels, which are composed of fuel pellets and fuel cladding, during an accident. To reduce the damage to the public, the fuels have to enhance their integrity under an accident environment. Enhanced accident tolerance fuels (ATFs) can tolerate a loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations as well as operational transients, in comparison with the current UO{sub 2}-Zr alloy system used in the LWR. Surface modified Zr cladding as a new concept was suggested to apply an enhanced ATF cladding. The aim of the partial ODS treatment is to increase the high-temperature strength to suppress the ballooning/rupture behavior of fuel cladding during an accident event. The target of the surface coating is to increase the corrosion resistance during normal operation and increase the oxidation resistance during an accident event. The partial ODS treatment of Zircaloy-4 cladding can be produced using a laser beam scanning method with Y2O3 powder, and the surface Cr-alloy and Cr/FeCrAl coating on Zircaloy-4 cladding can be obtained after the development of 3D laser coating and arc ion plating technologies.

  17. Pavement Snow Melting

    Energy Technology Data Exchange (ETDEWEB)

    Lund, John W.

    2005-01-01

    The design of pavement snow melting systems is presented based on criteria established by ASHRAE. The heating requirements depends on rate of snow fall, air temperature, relative humidity and wind velocity. Piping materials are either metal or plastic, however, due to corrosion problems, cross-linked polyethylene pipe is now generally used instead of iron. Geothermal energy is supplied to systems through the use of heat pipes, directly from circulating pipes, through a heat exchanger or by allowing water to flow directly over the pavement, by using solar thermal storage. Examples of systems in New Jersey, Wyoming, Virginia, Japan, Argentina, Switzerland and Oregon are presented. Key words: pavement snow melting, geothermal heating, heat pipes, solar storage, Wyoming, Virginia, Japan, Argentina, Klamath Falls.

  18. Ignition Delay Studies on Hypergolic Fuel Grains

    Directory of Open Access Journals (Sweden)

    S. R. Jain

    1988-07-01

    Full Text Available The ignition delays of several solid hypergolic fuel compositions, casted using various polymeric binders, or as melts, have been determined with fuming nitric acid as oxidizer. The ignition delays of various hypergolic fuel compositions increase drasticaliy on casting with binders like. carboxyl or hydroxyl termninated polybutadiene. Fuel grains cast using some newly syhthesised epoxy  resins with other ingrcdients, such as curing agent, magnesium powder and fuel, have short ignition delays of the order of 200 ms, and also good mechanical strength. Increasing the amount of binder in the composition retards the hypergolicity of the rain. Similar studies have been made on melt-cast systems using low melting hypergolic fuels for casting fuel powders. The ignition delays of the melt-cast grains, are longer than those determined taking the composition in the powder form. The effect of highly hypergolic additives, and metal powders, on the ignition delay of the cast compositions has been determined. Grains having good mechanical strength and short ignition delays have been obtained by optimising the fuel grain composition.

  19. Late-phase melt progression experiment: MP-2. Results and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gasser, R.D.; Gauntt, R.O.; Bourcier, S.C. [and others

    1997-05-01

    In-pile experiments addressing late-phase processes in Light Water Reactors (LWRs) were performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. Melt Progression (MP) experiments were designed to provide information to develop and verify computer models for analysis of LWR core damage in severe accidents. Experiments examine the formation and motion of ceramic molten pools in disrupted reactor core regions. The MP-2 experiment assembly consisted of: (1) a rubble bed of enriched UO{sub 2} and ZrO{sub 2} simulating severely disrupted reactor core regions, (2) a ceramic/metallic crust representing blockage formed by early phase melting, relocation, and refreezing of core components, and (3) an intact rod stub region that remained in place below the blockage region. The test assembly was fission heated in the central cavity of the ACRR at an average rate of about 0.2 KA, reaching a peak molten pool temperature around 3400 K. Melting of the debris bed ceramic components was initiated near the center of the bed. The molten material relocated downward, refreezing to form a ceramic crust near the bottom of the rubble bed. As power levels were increased, the crust gradually remelted and reformed at progressively lower positions in the bed until late in the experiment when it penetrated into and attacked the ceramic/metallic blockage. The metallic components of the blockage region melted and relocated to the bottom of the intact rod stub region before the ceramic melt penetrated the blockage region from above. The ceramic pool penetrated halfway into the blockage region by the end of the experiment. Measurements of thermal response and material relocation are compared to the results of the computer simulations. Postexperiment examination of the assembly with the associated material interactions and metallurgy are also discussed in detail with the analyses and interpretation of results. 16 refs., 206 figs., 24 tabs.

  20. Coated Particle Fuel and Deep Burn Program Monthly Highlights March 2011

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Lance Lewis [ORNL; Bell, Gary L [ORNL; Besmann, Theodore M [ORNL

    2011-04-01

    During FY 2011 the CP & DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for February 2011, ORNL/TM-2011/71, was distributed to program participants on March 8, 2011. As reported previously, the final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Thermochemical Data and Model Development - (a) Thermochemical Modeling, (b) Thermomechanical Behavior, (c) Actinide and Fission Product Transport, (d) Radiation Damage and Properties; (2) TRU (transuranic elements) TRISO (tri-structural isotropic) Development - (a) TRU Kernel Development, (b) Coating Development; (3) Advanced TRISO Applications - Metal Matrix Fuels for LWR; (4) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing; and (5) Fuel Performance and Analytical Analysis - Fuel Performance Modeling.

  1. Coated Particle Fuel and Deep Burn Program Monthly Highlights February 2011

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Lance Lewis [ORNL; Bell, Gary L [ORNL; Besmann, Theodore M [ORNL

    2011-03-01

    During FY 2011 the CP & DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for January 2010, ORNL/TM-2011/30, was distributed to program participants on February 8, 2011. As reported previously, the final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Thermochemical Data and Model Development - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (2) TRU (transuranic elements) TRISO (tri-structural isotropic) Development - (a) TRU Kernel Development, (b) Coating Development; (3) Advanced TRISO Applications - Metal Matrix Fuels for LWR; (4) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing; and (5) Fuel Performance and Analytical Analysis - Fuel Performance Modeling.

  2. Coated Particle Fuel and Deep Burn Program Monthly Highlights January 2011

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Lance Lewis [ORNL; Bell, Gary L [ORNL; Besmann, Theodore M [ORNL

    2011-02-01

    During FY 2011 the CP & DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for December 2010, ORNL/TM-2011/10, was distributed to program participants on January 12, 2011. As reported last month, the final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Thermochemical Data and Model Development - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (2) TRU (transuranic elements) TRISO (tri-structural isotropic) Development - (a) TRU Kernel Development, (b) Coating Development; (3) Advanced TRISO Applications - Metal Matrix Fuels for LWR; (4) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing; (5) Fuel Performance and Analytical Analysis - Fuel Performance Modeling.

  3. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    Energy Technology Data Exchange (ETDEWEB)

    G. S. Chang; Hongbin Zhang

    2009-09-01

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

  4. Standard guide for characterization of spent nuclear fuel in support of geologic repository disposal

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This guide provides guidance for the types and extent of testing that would be involved in characterizing the physical and chemical nature of spent nuclear fuel (SNF) in support of its interim storage, transport, and disposal in a geologic repository. This guide applies primarily to commercial light water reactor (LWR) spent fuel and spent fuel from weapons production, although the individual tests/analyses may be used as applicable to other spent fuels such as those from research and test reactors. The testing is designed to provide information that supports the design, safety analysis, and performance assessment of a geologic repository for the ultimate disposal of the SNF. 1.2 The testing described includes characterization of such physical attributes as physical appearance, weight, density, shape/geometry, degree, and type of SNF cladding damage. The testing described also includes the measurement/examination of such chemical attributes as radionuclide content, microstructure, and corrosion product c...

  5. Jules Horowitz Reactor Project- Fuel irradiation device, innovative instrumentation proposal for experimental phenomena real time measurement

    Energy Technology Data Exchange (ETDEWEB)

    Gaillot, Stephane; Cheymol, Guy [CEA, Paris (France)

    2013-07-01

    The fuel irradiation devices used for the tests or rods allow reproducing at small scales the conditions of the studied nuclear reactors (as LWR type). During the irradiation phase, the tested fuel rod can be stressed due to thermal, mechanical, nuclear effects which can modify its geometry (dilatation, swelling effects). After the test, the return to normal conditions can have as consequence the disappearance of the phenomenon or give access to partial information (final deformation). Generally, to follow the phenomena related to the irradiation phase, the experimental rod contained in the test device is instrumented with thermocouples and LVDT. As complement of this instrumentation, new sensors using innovating technologies are studied (deformation sensor integrating optical fibres). Through the example of a fuel irradiation device foreseen for the JHR, this paper aims to describe the present development of an innovating instrumentation with the objective to measure, in real time and under flux, the fuel rod deformation phenomena during a ramp test.

  6. Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen

    2012-09-01

    The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

  7. Mobile Melt-Dilute Technology Development Project FY 2005 Test Report

    Energy Technology Data Exchange (ETDEWEB)

    David A. Sell; Donald Fisher

    2006-01-01

    The adaptation of Melt-Dilute technology to a mobile and deployable platform progressed with the installation of the prototype air-cooled induction furnace and power generator in an ISO cargo container. Process equipment tests were conducted in FY’05 on two fronts: the melt container and its associated hardware and the mobile furnace and generator. Container design was validated through tests at elevated temperature and pressure, under vacuum, and subjected to impact. The Mobile Melt-Dilute (MMD) furnace and power source tests were completed per the plan. The tests provided information necessary to successfully melt and dilute HEU research reactor fuel assemblies.

  8. Study of irradiation induced restructuring of high burnup fuel - Use of computer and accelerator for fuel science and engineering -

    Energy Technology Data Exchange (ETDEWEB)

    Sataka, M.; Ishikawa, N.; Chimn, Y.; Nakamura, J.; Amaya, M. [Japan Atomic Energy Agency, Naka Gun (Japan); Iwasawa, M.; Ohnuma, T.; Sonoda, T. [Central Research Institute of Electric Power Industry, Tokyo (Japan); Kinoshita, M.; Geng, H. Y.; Chen, Y.; Kaneta, Y. [The Univ. of Tokyo, Tokyo (Japan); Yasunaga, K.; Matsumura, S.; Yasuda, K. [Kyushu Univ., Motooka (Japan); Iwase [Osaka Prefecture Univ., Osaka (Japan); Ichinomiya, T.; Nishiuran, Y. [Hokkaido Univ., Kitaku (Japan); Matzke, HJ. [Academy of Ceramics, Karlsruhe (Germany)

    2008-10-15

    In order to develop advanced fuel for future LWR reactors, trials were made to simulate the high burnup restructuring of the ceramics fuel, using accelerator irradiation out of pile and with computer simulation. The target is to reproduce the principal complex process as a whole. The reproduction of the grain subdivision (sub grain formation) was successful at experiments with sequential combined irradiation. It was made by recovery process of the accumulated dislocations, making cells and sub-boundaries at grain boundaries and pore surfaces. Details of the grain sub division mechanism is now in front of us outside of the reactor. Extensive computer science studies, first principle and molecular dynamics gave behavior of fission gas atoms and interstitial oxygen, assisting the high burnup restructuring.

  9. Uranium and neodymium partitioning in alkali chloride melts using low-melting gallium-based alloys

    Directory of Open Access Journals (Sweden)

    Melchakov Stanislav Yu.

    2015-12-01

    Full Text Available Partitioning of uranium and neodymium was studied in a ‘molten chloride salt - liquid Ga-X (X = In or Sn alloy’ system. Chloride melts were based on the low-melting ternary LiCl-KCl-CsCl eutectic. Nd/U separation factors were calculated from the thermodynamic data as well as determined experimentally. Separation of uranium and neodymium was studied using reductive extraction with neodymium acting as a reducing agent. Efficient partitioning of lanthanides (Nd and actinides (U, simulating fission products and fissile materials in irradiated nuclear fuels, was achieved in a single stage process. The experimentally observed Nd/U separation factor valued up to 106, depending on the conditions.

  10. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  11. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  12. Thermal Conductivity Measurement and Analysis of Fully Ceramic Microencapsulated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. G.; Kim, D. J.; Park, J. Y.; Kim, W. J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, S. J. [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of)

    2015-10-15

    FCM nuclear fuel is composed of tristructural isotropic(TRISO) fuel particle and SiC ceramic matrix. SiC ceramic matrix play an essential part in protecting fission product. In the FCM fuel concept, fission product is doubly protected by TRISO coating layer and SiC ceramic matrix in comparison with the current commercial UO2 fuel system of LWR. In addition to a safety enhancement of FCM fuel, thermal conductivity of SiC ceramic matrix is better than that of UO2 fuel. Because the centerline temperature of FCM fuel is lower than that of the current UO2 fuel due to the difference of thermal conductivity of fuel, an operational release of fission products from the fuel can be reduced. SiC ceramic has attracted for nuclear fuel application due to its high thermal conductivity properties with good radiation tolerant properties, a low neutron absorption cross-section and a high corrosion resistance. Thermal conductivity of ceramic matrix composite depends on the thermal conductivity of each component and the morphology of reinforcement materials such as fibers and particles. There are many results about thermal conductivity of fiber-reinforced composite like as SiCf/SiC composite. Thermal conductivity of SiC ceramics and FCM pellets with the volume fraction of TRISO particles were measured and analyzed by analytical models. Polycrystalline SiC ceramics and FCM pellets with TRISO particles were fabricated by hot press sintering with sintering additives. Thermal conductivity of the FCM pellets with TRISO particles of 0 vol.%, 10 vol.%, 20 vol.%, 30 vol.% and 40 vol.% show 68.4, 52.3, 46.8, 43.0 and 34.5 W/mK, respectively. As the volume fraction of TRISO particles increased, the measured thermal conductivity values closely followed the prediction of Maxwell's equation.

  13. Silicon carbide composite for light water reactor fuel assembly applications

    Science.gov (United States)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  14. Silicon carbide composite for light water reactor fuel assembly applications

    Energy Technology Data Exchange (ETDEWEB)

    Yueh, Ken, E-mail: kyueh@epri.com [Fuel Reliability Program, EPRI, 1300 West WT Harris Blvd, Charlotte, NC 28262 (United States); Terrani, Kurt A., E-mail: terranika@ornl.gov [Fusion and Materials for Nuclear Systems Division, Oak Ridge National Laboratory, 1 Bethel Valley Rd. MS 6093, Oak Ridge, TN 37831 (United States)

    2014-05-01

    The feasibility of using SiC{sub f}–SiC{sub m} composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  15. FUEL CONTAINER CLOSURE

    Science.gov (United States)

    Stone, C.C.; Noland, R.A.

    1962-07-17

    A method of producing a spherical closed end on a small thin-walled tube forming a container for reactor fuel is described. An end cap inserted in the tube has a long slender external projection extending axially of the tube. A heavy electrical current is applied to the projection, and simultaneously a portion of the tube very near the end cap is chilled. A part of the projection is vaporized, and the balance is melted, along with portions of the end cap and the tube. As a result, the end cap is welded to the tube, and the projection in melting is spread out to form a spherical surface on the end cap. (AEC)

  16. Fabrication and Characterization of UN-USix Nuclear Fuel

    OpenAIRE

    Raftery, Alicia Marie

    2015-01-01

    In this thesis, UN-U3Si2 nuclear fuel was fabricated using spark plasma sintering and characterized to analyze the microstructure and crystal structure of the resulting pellets. This work was done in collaboration with accident tolerant fuel research, an effort which aims at developing nuclear fuel with superior safety and performance compared to currently used oxide fuels. Uranium silicide was manufactured by arc melting to produce U3Si2 and uranium mononitride was synthesized by using the h...

  17. Propagation of neutron-reaction uncertainties through multi-physics models of novel LWR's

    Directory of Open Access Journals (Sweden)

    Hernandez-Solis Augusto

    2017-01-01

    Full Text Available The novel design of the renewable boiling water reactor (RBWR allows a breeding ratio greater than unity and thus, it aims at providing for a self-sustained fuel cycle. The neutron reactions that compose the different microscopic cross-sections and angular distributions are uncertain, so when they are employed in the determination of the spatial distribution of the neutron flux in a nuclear reactor, a methodology should be employed to account for these associated uncertainties. In this work, the Total Monte Carlo (TMC method is used to propagate the different neutron-reactions (as well as angular distributions covariances that are part of the TENDL-2014 nuclear data (ND library. The main objective is to propagate them through coupled neutronic and thermal-hydraulic models in order to assess the uncertainty of important safety parameters related to multi-physics, such as peak cladding temperature along the axial direction of an RBWR fuel assembly. The objective of this study is to quantify the impact that ND covariances of important nuclides such as U-235, U-238, Pu-239 and the thermal scattering of hydrogen in H2O have in the deterministic safety analysis of novel nuclear reactors designs.

  18. HTGR Technology Family Assessment for a Range of Fuel Cycle Missions

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Samuel E. Bays; Nick Soelberg

    2010-08-01

    This report examines how the HTGR technology family can provide options for the once through, modified open cycle (MOC), or full recycle fuel cycle strategies. The HTGR can serve all the fuel cycle missions that an LWR can; both are thermal reactors. Additional analyses are warranted to determine if HTGR “full recycle” service could provide improved consumption of transuranic (TRU) material than LWRs (as expected), to analyze the unique proliferation resistance issues associated with the “pebble bed” approach, and to further test and analyze methods to separate TRISO-coated fuel particles from graphite and/or to separate used HTGR fuel meat from its TRISO coating. The feasibility of these two separation issues is not in doubt, but further R&D could clarify and reduce the cost and enable options not adequately explored at present. The analyses here and the now-demonstrated higher fuel burnup tests (after the illustrative designs studied here) should enable future MOC and full recycle HTGR concepts to more rapidly consume TRU, thereby offering waste management advantages. Interest in “limited separation” or “minimum fuel treatment” separation approaches motivates study of impurity-tolerant fuel fabrication. Several issues are outside the scope of this report, including the following: thorium fuel cycles, gas-cooled fast reactors, the reliability of TRISO-coated particles (billions in a reactor), and how soon any new reactor or fuel type could be licensed and then deployed and therefore impact fuel cycle performance measures.

  19. Breeder Spent Fuel Handling (BSFH) cask study for FY83. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Diggs, J M

    1985-01-01

    This report documents a study conducted to investigate the applicability of existing LWR casks to shipment of long-cooled LMFBR fuel from the Clinch River Breeder Reactor Plant (CRBRP) to the Breeder Reprocessing Engineering Test (BRET) Facility. This study considered a base case of physical constraints of plants and casks, handling capabilities of plants, through-put requirements, shielding requirements due to transportation regulation, and heat transfer capabilities of the cask designs. Each cask design was measured relative to the base case. 15 references, 4 figures, 6 tables.

  20. Direct Coal Oxidation in Modified Solid Oxide Fuel Cells

    DEFF Research Database (Denmark)

    Deleebeeck, Lisa; Gil, Vanesa; Ippolito, Davide

    2017-01-01

    Hybrid direct carbon fuel cells employ a classical solid oxide fuel cell together with carbon dispersed in a carbonate melt on the anode side. In a European project, the utilization of various coals has been investigated with and without addition of an oxidation catalyst to the carbon-carbonate s......Hybrid direct carbon fuel cells employ a classical solid oxide fuel cell together with carbon dispersed in a carbonate melt on the anode side. In a European project, the utilization of various coals has been investigated with and without addition of an oxidation catalyst to the carbon...

  1. Melting of Ice under Pressure

    Energy Technology Data Exchange (ETDEWEB)

    Schwegler, E; Sharma, M; Gygi, F; Galli, G

    2008-07-31

    The melting of ice under pressure is investigated with a series of first principles molecular dynamics simulations. In particular, a two-phase approach is used to determine the melting temperature of the ice-VII phase in the range of 10 to 50 GPa. Our computed melting temperatures are consistent with existing diamond anvil cell experiments. We find that for pressures between 10 to 40 GPa, ice melts as a molecular solid. For pressures above {approx}45 GPa there is a sharp increase in the slope of the melting curve due to the presence of molecular dissociation and proton diffusion in the solid, prior to melting. The onset of significant proton diffusion in ice-VII as a function of increasing temperature is found to be gradual and bears many similarities to that of a type-II superionic solid.

  2. Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium–Plutonium Carbide based fuel and 300 MWt Power

    Science.gov (United States)

    Clief Pattipawaej, Sandro; Su’ud, Zaki

    2017-01-01

    A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

  3. LWRS Fuels Pathway: Engineering Design and Fuels Pathway Initial Testing of the Hot Water Corrosion System

    Energy Technology Data Exchange (ETDEWEB)

    Dr. John Garnier; Dr. Kevin McHugh

    2012-09-01

    The Advanced LWR Nuclear Fuel Development R&D pathway performs strategic research focused on cladding designs leading to improved reactor core economics and safety margins. The research performed is to demonstrate the nuclear fuel technology advancements while satisfying safety and regulatory limits. These goals are met through rigorous testing and analysis. The nuclear fuel technology developed will assist in moving existing nuclear fuel technology to an improved level that would not be practical by industry acting independently. Strategic mission goals are to improve the scientific knowledge basis for understanding and predicting fundamental nuclear fuel and cladding performance in nuclear power plants, and to apply this information in the development of high-performance, high burn-up fuels. These will result in improved safety, cladding, integrity, and nuclear fuel cycle economics. To achieve these goals various methods for non-irradiated characterization testing of advanced cladding systems are needed. One such new test system is the Hot Water Corrosion System (HWCS) designed to develop new data for cladding performance assessment and material behavior under simulated off-normal reactor conditions. The HWCS is capable of exposing prototype rodlets to heated, high velocity water at elevated pressure for long periods of time (days, weeks, months). Water chemistry (dissolved oxygen, conductivity and pH) is continuously monitored. In addition, internal rodlet heaters inserted into cladding tubes are used to evaluate repeated thermal stressing and heat transfer characteristics of the prototype rodlets. In summary, the HWCS provides rapid ex-reactor evaluation of cladding designs in normal (flowing hot water) and off-normal (induced cladding stress), enabling engineering and manufacturing improvements to cladding designs before initiation of the more expensive and time consuming in-reactor irradiation testing.

  4. Viscosity Meaurement Technique for Metal Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Ban, Heng [Utah State Univ., Logan, UT (United States). Mechanical and Aerospace Engineering; Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-09

    Metallic fuels have exceptional transient behavior, excellent thermal conductivity, and a more straightforward reprocessing path, which does not separate out pure plutonium from the process stream. Fabrication of fuel containing minor actinides and rare earth (RE) elements for irradiation tests, for instance, U-20Pu-3Am-2Np-1.0RE-15Zr samples at the Idaho National Laboratory, is generally done by melt casting in an inert atmosphere. For the design of a casting system and further scale up development, computational modeling of the casting process is needed to provide information on melt flow and solidification for process optimization. Therefore, there is a need for melt viscosity data, the most important melt property that controls the melt flow. The goal of the project was to develop a measurement technique that uses fully sealed melt sample with no Americium vapor loss to determine the viscosity of metallic melts and at temperatures relevant to the casting process. The specific objectives of the project were to: develop mathematical models to establish the principle of the measurement method, design and build a viscosity measurement prototype system based on the established principle, and calibrate the system and quantify the uncertainty range. The result of the project indicates that the oscillation cup technique is applicable for melt viscosity measurement. Detailed mathematical models of innovative sample ampoule designs were developed to not only determine melt viscosity, but also melt density under certain designs. Measurement uncertainties were analyzed and quantified. The result of this project can be used as the initial step toward the eventual goal of establishing a viscosity measurement system for radioactive melts.

  5. MULTIPLE MELTING IN NYLON 1010

    Institute of Scientific and Technical Information of China (English)

    FU Shuren; CHEN Taoyung

    1983-01-01

    Multiple melting behavior of nylon 1010 has been investigated by using DSC instrument. Effects of partial scanning, partial scanning and annealing, heating rate, cooling rate and stepwise annealing on the melting curve were studied. Experimental results indicate that the sample undergoes a process of continuous melting and recrystallization during DSC scanning. Nylon 1010 contains a distribution of crystallites of different degrees of perfection which is strongly dependent on its previous thermal history. From the structural reorganization point of view, the origin of double and multiple peaks of the melting curve is explained.

  6. NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL ASSOCIATED WITH A CLOSED FUEL CYCLE

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Ebbinghaus, B.; Sleaford, Brad W.; Wallace, R. K.; Collins, Brian A.; Hase, Kevin R.; Robel, Martin; Jarvinen, G. D.; Bradley, Keith S.; Ireland, J. R.; Johnson, M. W.; Prichard, Andrew W.; Smith, Brian W.

    2010-06-11

    This paper examines the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the various processing steps required for a closed fuel cycle. This paper combines the results from earlier studies that examined the attractiveness of SNM associated with the processing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR with new results for the final, repeated burning of SNM in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). The results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of "attractiveness levels" that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, or undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed.

  7. Dry spent fuel storage with the MACSTOR system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. [Atomic Energy of Canada Ltd., Montreal, PQ (Canada). CANDU Operations

    1996-10-01

    Atomic Energy of Canada Limited (AECL), and Transnuclear Inc. (TNI) began in 1989 the development of the concrete spent fuel storage system, called MACSTOR (Modular Air-Cooled Canister STORage) for use with LWR spent fuel assemblies. It is a hybrid system which combines the operational economies of metal cask technology with the capital economies of concrete technology. The MACSTOR Module is a monolithic, shielded concrete vault structure that can accommodate up to 20 spent fuel canisters. Each canister typically holds up to 21 PWR or 44 BWR spent fuel assemblies with a nominal fuel burn up rate of 40,000 MWD/MTU and a 7 year minimum cooling period. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. Thus, the utility can be assured of both positive cooling of the fuel and verification of the integrity of the fuel confinement boundary. The structure is seismically designed and is capable of withstanding site design basis accident events. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. The MACSTOR system can economically address a wide range of storage capacity requirements. The modular concept allows for flexibility in determining each module`s capacity. Starting with 8 canisters, the capacity can be increased by increments of 4 up to 20 canisters. The MACSTOR system is also flexible in accommodating the various spent fuel types from such reactors as VVER-440, VVER-1000 and RBMK 1500. (J.P.N.)

  8. Study on the fuel cycle cost of gas turbine high temperature reactor (GTHTR300). Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Takei, Masanobu; Katanishi, Shoji; Nakata, Tetsuo; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Oda, Takefumi; Izumiya, Toru [Nuclear Fuel Industries, Ltd., Tokyo (Japan)

    2002-11-01

    In the basic design of gas turbine high temperature reactor (GTHTR300), reduction of the fuel cycle cost has a large benefit of improving overall plant economy. Then, fuel cycle cost was evaluated for GTHTR300. First, of fuel fabrication for high-temperature gas cooled reactor, since there was no actual experience with a commercial scale, a preliminary design for a fuel fabrication plant with annual processing of 7.7 ton-U sufficient four GTHTR300 was performed, and fuel fabrication cost was evaluated. Second, fuel cycle cost was evaluated based on the equilibrium cycle of GTHTR300. The factors which were considered in this cost evaluation include uranium price, conversion, enrichment, fabrication, storage of spent fuel, reprocessing, and waste disposal. The fuel cycle cost of GTHTR300 was estimated at about 1.07 yen/kWh. If the back-end cost of reprocessing and waste disposal is included and assumed to be nearly equivalent to LWR, the fuel cycle cost of GTHTR300 was estimated to be about 1.31 yen/kWh. Furthermore, the effects on fuel fabrication cost by such of fuel specification parameters as enrichment, the number of fuel types, and the layer thickness were considered. Even if the enrichment varies from 10 to 20%, the number of fuel types change from 1 to 4, the 1st layer thickness of fuel changes by 30 {mu}m, or the 2nd layer to the 4th layer thickness of fuel changes by 10 {mu}m, the impact on fuel fabrication cost was evaluated to be negligible. (author)

  9. Transmutation Performance Analysis for Inert Matrix Fuels in Light Water Reactors and Computational Neutronics Methods Capabilities at INL

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Pope; Samuel E. Bays; S. Piet; R. Ferrer; Mehdi Asgari; Benoit Forget

    2009-05-01

    The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF, thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multi-pass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed. These analyses provide comparison of direct thermal recycling of PuNpAmCmCf, PuNpAm, PuNp, and Pu. The results of this infinite lattice assembly-wise study using SCALE 5.1 indicate that it may be feasible to recycle TRU in this manner using an otherwise typical PWR assembly without violating peaking factor limits.

  10. Current Trend in Furnace Technology in the Melting Industries

    Directory of Open Access Journals (Sweden)

    O.A. Ighodalo

    2011-06-01

    Full Text Available The aim of this study is to presents some of the current trend in Furnace technology as it pertains to the melting industries. Furnaces are applied in various industries for material processing. Large amounts of energy are usually consumed in the melting industries. The current trend in furnace technology is towards energy conservation, enhanced efficiency and productivity. The reduction of pollutant emissions are also taken into consideration due to their environmental impact. The various materials and processes for furnaces are discussed. The various strategies being employed towards furnace energy conservation, efficiency and productivity, and reduction of pollutant emissions are also discussed. Such strategies include the use of better fuel types such as natural gas; improved insulation and refractory materials; advanced burner designs such as high velocity, regenerative and recuperative burners; new combustion technologies such as air and fuel staging, flue gas recirculation technique. Mathematical modeling is also being employed for analysis and design purpose.

  11. LIFE Materials: Fuel Cycle and Repository Volume 11

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, H; Blink, J A

    2008-12-12

    The fusion-fission LIFE engine concept provides a path to a sustainable energy future based on safe, carbon-free nuclear power with minimal nuclear waste. The LIFE design ultimately offers many advantages over current and proposed nuclear energy technologies, and could well lead to a true worldwide nuclear energy renaissance. When compared with existing and other proposed future nuclear reactor designs, the LIFE engine exceeds alternatives in the most important measures of proliferation resistance and waste minimization. The engine needs no refueling during its lifetime. It requires no removal of fuel or fissile material generated in the LIFE engine. It leaves no weapons-attractive material at the end of life. Although there is certainly a need for additional work, all indications are that the 'back end' of the fuel cycle does not to raise any 'showstopper' issues for LIFE. Indeed, the LIFE concept has numerous benefits: (1) Per unit of electricity generated, LIFE engines would generate 20-30 times less waste (in terms of mass of heavy metal) requiring disposal in a HLW repository than does the current once-through fuel cycle. (2) Although there may be advanced fuel cycles that can compete with LIFE's low mass flow of heavy metal, all such systems require reprocessing, with attendant proliferation concerns; LIFE engines can do this without enrichment or reprocessing. Moreover, none of the advanced fuel cycles can match the low transuranic content of LIFE waste. (3) The specific thermal power of LIFE waste is initially higher than that of spent LWR fuel. Nevertheless, this higher thermal load can be managed using appropriate engineering features during an interim storage period, and could be accommodated in a Yucca-Mountain-like repository by appropriate 'staging' of the emplacement of waste packages during the operational period of the repository. The planned ventilation rates for Yucca Mountain would be sufficient for LIFE waste

  12. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    Energy Technology Data Exchange (ETDEWEB)

    Lindley, Benjamin A.; Parks, Geoffrey T. [University of Cambridge, Cambridge (United Kingdom); Franceschini, Fausto [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  13. Melt pool dynamics during selective electron beam melting

    Science.gov (United States)

    Scharowsky, T.; Osmanlic, F.; Singer, R. F.; Körner, C.

    2014-03-01

    Electron beam melting is a promising additive manufacturing technique for metal parts. Nevertheless, the process is still poorly understood making further investigations indispensable to allow a prediction of the part's quality. To improve the understanding of the process especially the beam powder interaction, process observation at the relevant time scale is necessary. Due to the difficult accessibility of the building area, the high temperatures, radiation and the very high scanning speeds during the melting process the observation requires an augmented effort in the observation equipment. A high speed camera in combination with an illumination laser, band pass filter and mirror system is suitable for the observation of the electron beam melting process. The equipment allows to observe the melting process with a high spatial and temporal resolution. In this paper the adjustment of the equipment and results of the lifetime and the oscillation frequencies of the melt pool for a simple geometry are presented.

  14. Advanced coal-fired glass melting development program

    Energy Technology Data Exchange (ETDEWEB)

    1991-05-01

    The objective of Phase 1 of the current contract was to verify the technical feasibility and economic benefits of Vortec's advanced combustion/melting technology using coal as the fuel of choice. The objective of the Phase 2 effort was to improve the performance of the primary components and demonstrate the effective operation of a subscale process heater system integrated with a glass separator/reservoir. (VC)

  15. Regulatory instrument review: Aging management of LWR cables, containment and basemat, reactor coolant pumps, and motor-operated valves

    Energy Technology Data Exchange (ETDEWEB)

    Werry, E.V.; Somasundaram, S.

    1995-09-01

    The results of Stage 2 of the Regulatory Instrument Review are presented in this volume. Selected regulatory instruments, such as the Code of Federal Regulations (CFR), US Nuclear Regulatory Commission (NRC), Regulatory Guides, and ASME Codes, were investigated to determine the extent to which these regulations apply aging management to selected safety-related components in nuclear power plants. The Regulatory Instrument Review was funded by the NRC under the Nuclear Plant Aging Research (NPAR) program. Stage 2 of the review focused on four safety-related structures and components; namely, cables, containment and basemat, reactor coolant pumps, and motor-operated valves. The review suggests that the primary-emphasis of the regulatory instruments was on the design, construction, start-up, and operation of a nuclear power plant, and that aging issues were primarily addressed after an aging-related problem was recognized. This Stage 2 review confirms the results of the prior review; (see Regulatory Instrument Review: Management of Aging of LWR Major Safety-Related Components NUREG/CR-5490. The observations indicate that the regulations generally address management of age-related degradation indirectly. Specific age-related degradation phenomena frequently are dealt with in bulletins and notices or through generic issues, letters, etc. The major recommendation of this report, therefore, is that the regulatory instruments should more directly and explicitly address the aging phenomenon and the management of the age-related degradation process.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Beyond the Melting Pot Reconsidered.

    Science.gov (United States)

    Anderson, Elijah

    2000-01-01

    Discusses the 1963 book, "Beyond the Melting Pot," which suggested that eventually the problem of different ethnicities in the U.S. would be resolved and society would become one melting pot. Examines how changes in immigration and economic structures have affected the issue, noting the devastating effect of the dominant culture's…

  19. Transient Testing of Nuclear Fuels and Materials in the United States

    Science.gov (United States)

    Wachs, Daniel M.

    2012-12-01

    The United States has established that transient irradiation testing is needed to support advanced light water reactors fuel development. The U.S. Department of Energy (DOE) has initiated an effort to reestablish this capability. Restart of the Transient Testing Reactor (TREAT) facility located at the Idaho National Laboratory (INL) is being considered for this purpose. This effort would also include the development of specialized test vehicles to support stagnant capsule and flowing loop tests as well as the enhancement of postirradiation examination capabilities and remote device assembly capabilities at the Hot Fuel Examination Facility. It is anticipated that the capability will be available to support testing by 2018, as required to meet the DOE goals for the development of accident-tolerant LWR fuel designs.

  20. FRED fuel behaviour code: Main models and analysis of Halden IFA-503.2 tests

    Energy Technology Data Exchange (ETDEWEB)

    Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Shestopalov, A., E-mail: shest@dhtp.kiae.ru [RRC' Kurchatov Institute' , Kurchatov sq, 123182 Moscow (Russian Federation)

    2011-07-15

    Highlights: > We developed a new fuel rod behaviour code named FRED. > Main models and assumptions are described. > The code was checked using the IFA-503.2 tests performed at the Halden reactor. - Abstract: The FRED fuel rod code is being developed for thermal and mechanical simulation of fast breeder reactor (FBR) and light-water reactor (LWR) fuel behaviour under base-irradiation and accident conditions. The current version of the code calculates temperature distribution in fuel rods, stress-strain condition of cladding, fuel deformation, fuel-cladding gap conductance, and fuel rod inner pressure. The code was previously evaluated in the frame of two OECD mixed plutonium-uranium oxide (MOX) fuel performance benchmarks and then integrated into PSI's FAST code system to provide the fuel rod temperatures necessary for the neutron kinetics and thermal-hydraulic modules in transient calculations. This paper briefly overviews basic models and material property database of the FRED code used to assess the fuel behaviour under steady-state conditions. In addition, the code was used to simulate the IFA-503.2 tests, performed at the Halden reactor for two PWR and twelve VVER fuel samples under base-irradiation conditions. This paper presents the results of this simulation for two cases using a code-to-data comparison of fuel centreline temperatures, internal gas pressures, and fuel elongations. This comparison has demonstrated that the code adequately describes the important physical mechanisms of the uranium oxide (UOX) fuel rod thermal performance under steady-state conditions. Future activity should be concentrated on improving the model and extending the validation range, especially to the MOX fuel steady-state and transient behaviour.

  1. Assessment of Startup Fuel Options for a Test or Demonstration Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, Jon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Walters, L. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This document explores startup fuel options for a proposed test/demonstration fast reactor. The fuel options considered are the metallic fuels U-Zr and U-Pu-Zr and the ceramic fuels UO2 and UO2-PuO2 (MOX). Attributes of the candidate fuel choices considered were feedstock availability, fabrication feasibility, rough order of magnitude cost and schedule, and the existing irradiation performance database. The reactor-grade plutonium bearing fuels (U-Pu-Zr and MOX) were eliminated from consideration as the initial startup fuels because the availability and isotopics of domestic plutonium feedstock is uncertain. There are international sources of reactor grade plutonium feedstock but isotopics and availability are also uncertain. Weapons grade plutonium is the only possible source of Pu feedstock in sufficient quantities needed to fuel a startup core. Currently, the available U.S. source of (excess) weapons-grade plutonium is designated for irradiation in commercial light water reactors (LWR) to a level that would preclude diversion. Weapons-grade plutonium also contains a significant concentration of gallium. Gallium presents a potential issue for both the fabrication of MOX fuel as well as possible performance issues for metallic fuel. Also, the construction of a fuel fabrication line for plutonium fuels, with or without a line to remove gallium, is expected to be considerably more expensive than for uranium fuels. In the case of U-Pu-Zr, a relatively small number of fuel pins have been irradiated to high burnup, and in no case has a full assembly been irradiated to high burnup without disassembly and re-constitution. For MOX fuel, the irradiation database from the Fast Flux Test Facility (FFTF) is extensive. If a significant source of either weapons-grade or reactor-grade Pu became available (i.e., from an international source), a startup core based on Pu could be reconsidered.

  2. Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States); Charlton, William S., E-mail: wcharlton@tamu.edu [Nuclear Security Science and Policy Institute, Texas A and M University, 3473 TAMU, College Station, TX 77843 (United States); Menlove, Howard O., E-mail: hmenlove@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States); Swinhoe, Martyn T., E-mail: swinhoe@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States)

    2012-07-11

    A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four {sup 235}U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from {sup 235}U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15 Multiplication-Sign 15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective {sup 235}U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies. - Highlights: Black-Right-Pointing-Pointer Development of new measurement technique called SINRD to improve LWR safeguards. Black-Right-Pointing-Pointer Performed SINRD experiment to measure {sup 235}U and pin diversions in PWR fresh assembly. Black-Right-Pointing-Pointer Excellent agreement of MCNPX and measured results confirmed accuracy of SINRD model. Black-Right-Pointing-Pointer SINRD

  3. Fault rheology beyond frictional melting.

    Science.gov (United States)

    Lavallée, Yan; Hirose, Takehiro; Kendrick, Jackie E; Hess, Kai-Uwe; Dingwell, Donald B

    2015-07-28

    During earthquakes, comminution and frictional heating both contribute to the dissipation of stored energy. With sufficient dissipative heating, melting processes can ensue, yielding the production of frictional melts or "pseudotachylytes." It is commonly assumed that the Newtonian viscosities of such melts control subsequent fault slip resistance. Rock melts, however, are viscoelastic bodies, and, at high strain rates, they exhibit evidence of a glass transition. Here, we present the results of high-velocity friction experiments on a well-characterized melt that demonstrate how slip in melt-bearing faults can be governed by brittle fragmentation phenomena encountered at the glass transition. Slip analysis using models that incorporate viscoelastic responses indicates that even in the presence of melt, slip persists in the solid state until sufficient heat is generated to reduce the viscosity and allow remobilization in the liquid state. Where a rock is present next to the melt, we note that wear of the crystalline wall rock by liquid fragmentation and agglutination also contributes to the brittle component of these experimentally generated pseudotachylytes. We conclude that in the case of pseudotachylyte generation during an earthquake, slip even beyond the onset of frictional melting is not controlled merely by viscosity but rather by an interplay of viscoelastic forces around the glass transition, which involves a response in the brittle/solid regime of these rock melts. We warn of the inadequacy of simple Newtonian viscous analyses and call for the application of more realistic rheological interpretation of pseudotachylyte-bearing fault systems in the evaluation and prediction of their slip dynamics.

  4. Fuel Performance Characterisation under Various PWR Conditions: Description of the Annealing Test Facilities available at the LECA-STAR laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Pontillon, Y.; Cornu, B.; Clement, S.; Ferroud-Plattet, M.P.; Malgouyres, P.P. [Commissariat a l' Energie Atomique, CEA/DEN/DEC/SA3C - Centre d' Etudes de Cadarache, BP1, 13108 Saint Paul Lez Durance (France)

    2008-07-01

    The aim to improve LWR fuel behaviour has led Cea to improve its post-irradiation examination capacities in term of test facilities and characterization techniques in the shielded hot cells of the LECA-STAR facility, located in Cadarache Cea center. as far as the annealing test facilities are concerned, fuel qualification and improvement of knowledge require a set of furnaces which are already used or will be used. The main characteristics of these furnaces strongly depend on the experimental objectives. The aim of this paper is to review the main aspects of these specific experiments concerning: (i) fission gas release from high burn up fuel, (ii) global fission product release in severe-accident conditions and (iii) fuel microstructural changes, potential cladding failure, radionuclide source terms... under conditions representative of long term dry storage and geological disposal. (authors)

  5. Phase change predictions for liquid fuel in contact with steel structure using the heat conduction equation

    Energy Technology Data Exchange (ETDEWEB)

    Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    When liquid fuel makes contact with steel structure the liquid can freeze as a crust and the structure can melt at the surface. The melting and freezing processes that occur can influence the mode of fuel freezing and hence fuel relocation. Furthermore the temperature gradients established in the fuel and steel phases determine the rate at which heat is transferred from fuel to steel. In this memo the 1-D transient heat conduction equations are applied to the case of initially liquid UO{sub 2} brought into contact with solid steel using up-to-date materials properties. The solutions predict criteria for fuel crust formation and steel melting and provide a simple algorithm to determine the interface temperature when one or both of the materials is undergoing phase change. The predicted steel melting criterion is compared with available experimental results. (author)

  6. Utilizing Rice Husk Briquettes in Firing Crucible Furnace for Low Temperature Melting Metals in Nigeria

    Directory of Open Access Journals (Sweden)

    N. A. Musa

    2012-08-01

    Full Text Available The search for alternative fuels for firing crucible furnace for low temperature melting metals has become mandatory, as a result of the pollution problem associated with the use of fossil fuels, the expense of electricity and also deforestation as a result of the use of charcoal. An agricultural waste, rice husk, in briquette form was used as an alternative fuel to fire crucible furnace to melt lead, zinc and aluminium. Results showed that lead and zinc melted and reached their pouring temperatures of 3840C and 5300C in 70 minutes and 75 minutes respectively. Aluminium was raised to a maximum temperature of 5200C in 75 and 100 minutes.The average concentration of the pollutants (CO, SO2and NOX were found to be below the tolerance limit and that of TSP (Total Suspended Particulates was found to be within the tolerance limit stipulated by Federal Environmental Protection Agency (FEPA in Nigeria.

  7. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  8. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  9. Double-clad nuclear fuel safety rod

    Science.gov (United States)

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  10. Characterization and heading of irradiated fuels and their chemical analogs; Caracterizacion y lixiviacion de combustibles nucleares irradiados y de sus analogos quimicos

    Energy Technology Data Exchange (ETDEWEB)

    Serrano, J. A. [Ciemat.Madrid (Spain)

    2000-07-01

    This work presents results of leaching experiments under deionized water and under synthetic granite at room temperature in air using spent fuel (UO{sub 2} and MOX LWR fuels) and the chemical analogues, natural UO{sub 2} and SIMFUEL. The experimental conditions and procedure for irradiated and non-irradiated materials were kept similar as much as possible. Also dissolution behaviour studies of preoxidised LWR UO{sub 2} and MOX spent fuel up to different on the oxidation degree. For both fuel types, UO{sub 2} and MOX, the fission products considered showed a fractional release normalised to uranium higher than 1, due to either the larger inventory at preferential leaching zones, such as, grain boundaries or to the inherent higher solubility of some of these elements. In contrast to fission products, the fractional release of PU from the UO{sub 2} fuel was not affected by the oxidation level. Finally a thermodynamic study of the experimental leaching results obtained in this work was performed. (Author)

  11. Electrical Conductivity of Cryolite Melts

    Science.gov (United States)

    Fellner, P.; Grjotheim, K.; Kvande, H.

    1985-11-01

    This paper proposes an equation for the electrical conductivity of multicomponent cryolite-based mixtures. The equation is based on a physical model which assumes that the conductivity is proportional to the number density of the effective electric charges in the melt. The various authors in the available literature show a great discrepancy in conductivity data of cryolite-based melts. The equation based on the physical model enables determination of which set of data is preferable. Special consideration in this respect is given to the influence of magnesium flouride and lithium flouride additions to the melt.

  12. Fuel distribution

    Energy Technology Data Exchange (ETDEWEB)

    Tison, R.R.; Baker, N.R.; Blazek, C.F.

    1979-07-01

    Distribution of fuel is considered from a supply point to the secondary conversion sites and ultimate end users. All distribution is intracity with the maximum distance between the supply point and end-use site generally considered to be 15 mi. The fuels discussed are: coal or coal-like solids, methanol, No. 2 fuel oil, No. 6 fuel oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Although the fuel state, i.e., gas, liquid, etc., can have a major impact on the distribution system, the source of these fuels (e.g., naturally-occurring or coal-derived) does not. Single-source, single-termination point and single-source, multi-termination point systems for liquid, gaseous, and solid fuel distribution are considered. Transport modes and the fuels associated with each mode are: by truck - coal, methanol, No. 2 fuel oil, and No. 6 fuel oil; and by pipeline - coal, methane, No. 2 fuel oil, No. 6 oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Data provided for each distribution system include component makeup and initial costs.

  13. Revisiting the melting temperature of NpO2 and the challenges associated with high temperature actinide compound measurements

    NARCIS (Netherlands)

    Böhler, R.; Welland, M.J.; De Bruycker, F.; Boboridis, K.; Janssen, A.; Eloirdi, R.; Konings, R.J.M.; Manara, D.

    2012-01-01

    This work revisits the melting behaviour of neptunium dioxide, an actinide compound which can be produced in the nuclear fuel during operation, and which has an important impact on the nuclear fuel and waste radioactivity especially on the very long term. The present experimental approach employs re

  14. Direct Coal Oxidation in Modified Solid Oxide Fuel Cells

    DEFF Research Database (Denmark)

    Deleebeeck, Lisa; Gil, Vanesa; Ippolito, Davide;

    2015-01-01

    Hybrid direct carbon fuel cells employ a classical solid oxide fuel cell together with carbon dispersed in a carbonate melt on the anode side. In a European project, the utilization of various coals has been investigated with and without addition of an oxidation catalyst to the carbon-carbonate s...

  15. Nitrogen Control in VIM Melts

    Science.gov (United States)

    Jablonski, P. D.; Hawk, J. A.

    NETL has developed a design and control philosophy for the addition of nitrogen to austenitic and ferritic steels. The design approach uses CALPHAD as the centerpiece to predict the level to which nitrogen is soluble in both the melt and the solid. Applications of this technique have revealed regions of "exclusion" in which the alloy, while within specification limits of prescribed, cannot be made by conventional melt processing. Furthermore, other investigations have found that substantial retrograde solubility of nitrogen exists, which can become problematic during subsequent melt processing and/or other finishing operations such as welding. Additionally, the CALPHAD method has been used to adjust primary melt conditions. To that end, nitrogen additions have been made using chrome nitride, silicon nitride, high-nitrogen ferrochrome as well as nitrogen gas. The advantages and disadvantages of each approach will be discussed and NETL experience in this area will be summarized with respect to steel structure.

  16. Nondestructive, energy-dispersive, x-ray fluorescence analysis of actinide stream concentrations from reprocessed nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Camp, D.C.; Ruhter, W.D.

    1979-06-27

    In one plan for reprocessing LWR spent fuel, after separation from fission products and transplutonics, part of the U and all of the Pu in a nitrate solution will form a coprocessed stream which is then evaporated and sent to a hold tank for accounting. The remaining U fraction will be purified and sent to a separate storage tank. These two streams can be monitored using x-ray fluorescence analysis. This report discusses equipment, spectra, cell calibration, and dynamic concentration measurements. 7 figures. (DLC)

  17. TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    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

  18. Scaleable Clean Aluminum Melting Systems

    Energy Technology Data Exchange (ETDEWEB)

    Han, Q.; Das, S.K. (Secat, Inc.)

    2008-02-15

    The project entitled 'Scaleable Clean Aluminum Melting Systems' was a Cooperative Research and Development Agreements (CRADAs) between Oak Ridge National Laboratory (ORNL) and Secat Inc. The three-year project was initially funded for the first year and was then canceled due to funding cuts at the DOE headquarters. The limited funds allowed the research team to visit industrial sites and investigate the status of using immersion heaters for aluminum melting applications. Primary concepts were proposed on the design of furnaces using immersion heaters for melting. The proposed project can continue if the funding agency resumes the funds to this research. The objective of this project was to develop and demonstrate integrated, retrofitable technologies for clean melting systems for aluminum in both the Metal Casting and integrated aluminum processing industries. The scope focused on immersion heating coupled with metal circulation systems that provide significant opportunity for energy savings as well as reduction of melt loss in the form of dross. The project aimed at the development and integration of technologies that would enable significant reduction in the energy consumption and environmental impacts of melting aluminum through substitution of immersion heating for the conventional radiant burner methods used in reverberatory furnaces. Specifically, the program would couple heater improvements with furnace modeling that would enable cost-effective retrofits to a range of existing furnace sizes, reducing the economic barrier to application.

  19. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: reprocessing light-water reactor fuel. [Radiation dose commitment to human populations from radioactive effluents released to environment

    Energy Technology Data Exchange (ETDEWEB)

    Finney, B.C.; Blanco, R.E.; Dahlman, R.C.; Hill, G.S.; Kitts, F.G.; Moore, R.E.; Witherspoon, J.P.

    1976-10-01

    A cost/benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model nuclear fuel reprocessing plant which processes light-water reactor (LWR) fuels, and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term as low as reasonably achievable in relation to limiting the release of radioactive materials from nuclear facilities. The base case model plant is representative of current plant technology and has an annual capacity of 1500 metric tons of LWR fuel. Additional radwaste treatment systems are added to the base case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The cost for the added waste treatment operations and the corresponding dose commitments are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Much of the technology used in the advanced cases is in an early stage of development and is not suitable for immediate use. The methodology used in estimating the costs, and the radiological doses, detailed calculations, and tabulations are presented in Appendix A and ORNL-4992. This report is a revision of the original study (ORNL/TM-4901).

  20. Two dimensional structural analysis of reactor fuel element claddings due to local effects

    Energy Technology Data Exchange (ETDEWEB)

    Karimi, R; Wolf, L

    1978-04-01

    Two dimensional thermoelastic and inelastic stresses and deformation of typical LWR (PWR) and LMFBR (CRBR) claddings are evaluated by utilizing the following codes, for (1) Thermoelastic analysis (a) STRESS Code (b) SEGPIPE Code (2) Thermoinelastic analysis (a) Modified version of the GOGO code (b) One dimensional GRO-II code. The primary objective of this study is to analyze the effect of various local perturbations in the clad temperature field, namely eccentrically mounted fuel pellet, clad ovality, power tilt across the fuel and clad-coolant heat transfer variation on the cladding stress and deformation. In view of the fact that the thermoelastic analysis is always the first logical choice entering the structural field, it was decided to start the analysis with the two dimensional codes such as STRESS and SEGPIPE. Later, in order to assess the validity and compare the thermoelastic results to those obtained for actual reactor conditions, a two dimensional code, namely a modified version of the GOGO code, was used to account for inelastic effects such as irradiation and thermal creep and swelling in the evaluation. The comparison of thermoelastic and inelastic results shows that the former can be used effectively to analyze LWR fuel pin over 350 hours of lifetime under the most adverse condition and 500 hours of lifetime for an LMFBR fuel pin. Beyond that the inelastic solution must be used. The impact of the individual thermal perturbation and combinations thereof upon the structural quantity is also shown. Finally, the effect of rod displacement on the two dimensional thermal and structural quantities of the LMFBR fuel pin cladding is analyzed.

  1. Studies of the UO 2-zircaloy chemical interaction and fuel rod relocation modes in a severe fuel damage accident

    Science.gov (United States)

    Shiozawa, S.; Ichikawa, M.; Fujishiro, T.

    1988-06-01

    Experiments have been conducted in the Nuclear Safety Research Reactor (NSRR) at JAERI since 1975 in order to study fuel rod failure behavior under reactivity-initiated accident conditions. Recently the experiments have been focussed on fuel behavior under simulated severe fuel damage (SFD) accident conditions. UO 2-Zircaloy reaction kinetics during very rapid transients at elevated temperatures was studied from a metallurgical point of view. Equilibrium was found to be established even in very rapid transients. The reaction rate equations developed in isothermal studies can be applied to interpret the experimental results. A fuel rod relocation criterion in connection with peak temperatures, environment conditions and initial fuel rod conditions was developed. According to the test results, fuel rod melt down due to liquefaction seems unlikely below the melting temperature of β-Zircaloy.

  2. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Gauld, Ian C [ORNL

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  3. Heterogeneous fuel for hybrid rocket

    Science.gov (United States)

    Stickler, David B. (Inventor)

    1996-01-01

    Heterogeneous fuel compositions suitable for use in hybrid rocket engines and solid-fuel ramjet engines, The compositions include mixtures of a continuous phase, which forms a solid matrix, and a dispersed phase permanently distributed therein. The dispersed phase or the matrix vaporizes (or melts) and disperses into the gas flow much more rapidly than the other, creating depressions, voids and bumps within and on the surface of the remaining bulk material that continuously roughen its surface, This effect substantially enhances heat transfer from the combusting gas flow to the fuel surface, producing a correspondingly high burning rate, The dispersed phase may include solid particles, entrained liquid droplets, or gas-phase voids having dimensions roughly similar to the displacement scale height of the gas-flow boundary layer generated during combustion.

  4. Quantifying melting and mobilistaion of interstitial melts in crystal mushes

    Science.gov (United States)

    Veksler, Ilya; Dobson, Katherine; Hess, Kai-Uwe; Ertel-Ingrisch, Werner; Humphreys, Madeleine

    2015-04-01

    The deformation of crystals mushes and separation of melts and crystals in is critical to understanding the development of physical and chemical heterogeneity in magma chambers and has been invoked as an eruption trigger mechanism. Here we investigate the behaviour of the melt in the well characterised, classic crystal mush system of the Skaergaard intrusion by combining experimental petrology and the non-destructive 3D imaging methods. Starting materials for partial melting experiments were four samples from the upper Middle Zone of the Layered Series. Cylinders, 15 mm in diameter and 20 mm in length, were drilled out of the rock samples, placed in alumina crucibles and held for 5 days in electric furnaces at atmospheric pressure and 1050-1100 °C. Redox conditions set by the CO-CO2 gas mixture were kept close to those of the FMQ buffer. We then use spatially registered 3D x-ray computed tomography images, collected before and after the experiment, to determine the volume and distribution of the crystal framework and interstitial phases, and the volume, distribution and connectivity the interstitial phases that undergo melting and extraction while at elevated temperature. Image analysis has allowed us to quantify these physical changes with high spatial resolution. Our work is a first step towards quantitative understanding of the melt mobilisation and migration processes operating in notionally locked crystal rich magmatic systems.

  5. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

  6. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  7. Methods for Melting Temperature Calculation

    Science.gov (United States)

    Hong, Qi-Jun

    Melting temperature calculation has important applications in the theoretical study of phase diagrams and computational materials screenings. In this thesis, we present two new methods, i.e., the improved Widom's particle insertion method and the small-cell coexistence method, which we developed in order to capture melting temperatures both accurately and quickly. We propose a scheme that drastically improves the efficiency of Widom's particle insertion method by efficiently sampling cavities while calculating the integrals providing the chemical potentials of a physical system. This idea enables us to calculate chemical potentials of liquids directly from first-principles without the help of any reference system, which is necessary in the commonly used thermodynamic integration method. As an example, we apply our scheme, combined with the density functional formalism, to the calculation of the chemical potential of liquid copper. The calculated chemical potential is further used to locate the melting temperature. The calculated results closely agree with experiments. We propose the small-cell coexistence method based on the statistical analysis of small-size coexistence MD simulations. It eliminates the risk of a metastable superheated solid in the fast-heating method, while also significantly reducing the computer cost relative to the traditional large-scale coexistence method. Using empirical potentials, we validate the method and systematically study the finite-size effect on the calculated melting points. The method converges to the exact result in the limit of a large system size. An accuracy within 100 K in melting temperature is usually achieved when the simulation contains more than 100 atoms. DFT examples of Tantalum, high-pressure Sodium, and ionic material NaCl are shown to demonstrate the accuracy and flexibility of the method in its practical applications. The method serves as a promising approach for large-scale automated material screening in which

  8. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

    2007-02-07

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  9. Core burnup calculation and accidents analyses of a pressurized water reactor partially loaded with rock-like oxide fuel

    Science.gov (United States)

    Akie, H.; Sugo, Y.; Okawa, R.

    2003-06-01

    A rock-like oxide (ROX) fuel - light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO 2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the power peaking tends to be large in this heterogeneous core configuration. The reactivity initiated accident (RIA) and loss of coolant accident (LOCA) behaviors were not sufficiently improved. In order to reduce the power peaking, the fuel composition and the assembly design of the ROX fuel were modified. Firstly, erbium burnable poison was added as Er 2O 3 in the ROX fuel to reduce the burnup reactivity swing. Then pin-by-pin Pu enrichment and Er content distributions within the ROX fuel assembly were considered. In addition, the Er content distribution was also considered in the axial direction of the ROX fuel pin. With these modifications, a power peaking factor even lower than the one in a conventional UO 2 fueled core can be obtained. The RIA and LOCA analyses of the modified core have also shown the comparable transient behaviors of ROX partial loading core to those of the UO 2 core.

  10. Light water reactor fuel element suitable for thorium employment in a discrete seed and blanket configuration with the aim to attain conversion ratios above the range of one

    Energy Technology Data Exchange (ETDEWEB)

    Hrovat, M.F.; Grosse, K.H.; Seemann, R. [ALD Vacuum Technologies GmbH, Hanau (Germany)

    2008-07-01

    The thorium resources in the world are relatively large. According to the IAEA-NEA-publication ''Red Book'' they amount to 4.5 10E6 metric tons and are about 4 times greater than the resources of Uranium. The fuel element described in this paper could be used in light water reactor (LWR) preferably in pressurized water reactor (PWR). The seed (feed) rods contain uranium 235 as fissionable material and the blanket (breed) rods contain thorium and uranium. The thorium in the blanket rods is converted to fissionable U-233 by irradiation with thermal neutrons. The U-233 produced is a valuable fissionable material and is characterized by high revalues, where t is defined as the number of fission neutrons per absorption in fissile materials. By optimized configuration and loading of the seed- and blanket rods the thorium is converted to U-233 and the U-238 is converted to fissionable Plutonium isotopes. Consequently more fissionable material is generated than is used. The fuel cycle is also flexible. Thus U-235, Pu-239 or weapons-grade Plutonium can be used.Based on knowledge obtained in the development of fuel elements for material test reactors (MTR), high temperature reactors (HTR) and light water reactors (LWR), a new design of fuel element suitable for thorium employment in PWR is described.

  11. Melting of superheated molecular crystals

    Science.gov (United States)

    Cubeta, Ulyana; Bhattacharya, Deepanjan; Sadtchenko, Vlad

    2017-07-01

    Melting dynamics of micrometer scale, polycrystalline samples of isobutane, dimethyl ether, methyl benzene, and 2-propanol were investigated by fast scanning calorimetry. When films are superheated with rates in excess of 105 K s-1, the melting process follows zero-order, Arrhenius-like kinetics until approximately half of the sample has transformed. Such kinetics strongly imply that melting progresses into the bulk via a rapidly moving solid-liquid interface that is likely to originate at the sample's surface. Remarkably, the apparent activation energies for the phase transformation are large; all exceed the enthalpy of vaporization of each compound and some exceed it by an order of magnitude. In fact, we find that the crystalline melting kinetics are comparable to the kinetics of dielectric α-relaxation in deeply supercooled liquids. Based on these observations, we conclude that the rate of non-isothermal melting for superheated, low-molecular-weight crystals is limited by constituent diffusion into an abnormally dense, glass-like, non-crystalline phase.

  12. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.

  14. Coated Particle Fuel and Deep Burn Program Monthly Highlights April 2011

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Lance Lewis [ORNL; Bell, Gary L [ORNL; Besmann, Theodore M [ORNL

    2011-05-01

    The baseline change proposal BCP-FCRD-11026 submitted to change the due date for M21AF080202 'Demonstrate fabrication of Transuranic kernels of Plutonium-239/3.5at%Neptunium-237 using newly installed glove box facilities in ORNL 7930 hot cell complex' from 4/25/11 to 3/30/12 was approved this month. During FY 2011 the CP & DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for March 2011, ORNL/TM-2011/96, was distributed to program participants on April 8, 2011. As reported previously, the final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Thermochemical Data and Model Development - (a) Thermochemical Modeling, (b) Thermomechanical Behavior, (c) Actinide and Fission Product Transport, (d) Radiation Damage and Properties; (2) TRU (transuranic elements) TRISO (tri-structural isotropic) Development - (a) TRU Kernel Development, (b) Coating Development; (3) Advanced TRISO Applications - Metal Matrix Fuels for LWR; (4) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing; (5) Fuel Performance and Analytical Analysis - Fuel Performance Modeling; and (6) ZrC Properties and Handbook - Properties of ZrC.

  15. The use of computed neutron coincidence counting with time interval analysis for the analysis of Fork-measurements on a fresh MOX-LWR fuel assembly under water

    Energy Technology Data Exchange (ETDEWEB)

    Baeten, P.; Bruggeman, M.; Carchon, R

    1998-06-01

    The objective of this study was to investigate the influence of different important parameters on measurement results for various fork-detectors. Computed Neutron Coincidence Counting (CNCC) with Time Interval Analysis (TIA) was used for this study. The performance of the electronics for the different fork-detectors was studied by investigating the deadtime perturbed zone of the Rossi-alpha distribution in TIA. The measurement revealed anomalies in the performance of the electronics of the IAEA BWR and LANL fork-detector. The IAEA PWR fork-detector functioned well and the deadtime parameter was calculated. The optimal setting for the pre delay was investigated and it was found that a pre delay of 10 micro seconds should be considered as an optimum between excluding from analysis data in the deadtime perturbed zone and keeping a high signal-to-noise ratio. For the shift register electronics used with the fork-detectors, a pre delay of only 4.5 micro seconds was used. The study of the pre delay and the deadtime showed that the calculated triples-rate is strongly dependent on these parameters. An accurate determination of the triple-rate in this type of measurements has proven to be quite difficult and requires proper operation of the electronics, a correct pre delay and an accurate deadtime correction formalism. By varying the boron concentration in water, the change of the decay time of the Rossi-alpha distribution was clearly observed. This change is due to the variation of the thermal multiplication. The variation of this decay time with the boron concentration proves that Boehnel's model for fast neutron multiplication is not valid under these measurement conditions and that a model for fast and thermal multiplication should be used in order to obtain unbiased measurement results. CNCC with TIA has proved to be a valuable tool in which parameter settings can be varied a posterori and the optimal setting can be determined for each measurement. Moreover, the display of the time intervals allows a simple control of the proper system performance.

  16. ELCOS: the PSI code system for LWR core analysis. Part II: user`s manual for the fuel assembly code BOXER

    Energy Technology Data Exchange (ETDEWEB)

    Paratte, J.M.; Grimm, P.; Hollard, J.M. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-02-01

    ELCOS is a flexible code system for the stationary simulation of light water reactor cores. It consists of the four computer codes ETOBOX, BOXER, CORCOD and SILWER. The user`s manual of the second one is presented here. BOXER calculates the neutronics in cartesian geometry. The code can roughly be divided into four stages: - organisation: choice of the modules, file manipulations, reading and checking of input data, - fine group fluxes and condensation: one-dimensional calculation of fluxes and computation of the group constants of homogeneous materials and cells, - two-dimensional calculations: geometrically detailed simulation of the configuration in few energy groups, - burnup: evolution of the nuclide densities as a function of time. This manual shows all input commands which can be used while running the different modules of BOXER. (author) figs., tabs., refs.

  17. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 4. Alternatives for waste isolation and disposal

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Volume IV of the five-volume report contains information on alternatives for final storage and disposal of radioactive wastes. Section titles include: basic concepts for geologic isolation; geologic storage alternatives; geologic disposal alternatives; extraterrestrial disposal; and, transmutation. (JGB)

  18. Transition Analysis of Promising U.S. Future Fuel Cycles Using ORION

    Energy Technology Data Exchange (ETDEWEB)

    Sunny, Eva E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gregg, Robert [Preston Laboratory, Springfields Works (United Kingdom)

    2015-01-01

    The US Department of Energy Office of Fuel Cycle Technologies performed an evaluation and screening (E&S) study of nuclear fuel cycle options to help prioritize future research and development decisions. Previous work for this E&S study focused on establishing equilibrium conditions for analysis examples of 40 nuclear fuel cycle evaluation groups (EGs) and evaluating their performance according to a set of 22 standardized metrics. Following the E&S study, additional studies are being conducted to assess transitioning from the current US fuel cycle to future fuel cycle options identified by the E&S study as being most promising. These studies help inform decisions on how to effectively achieve full transition, estimate the length of time needed to undergo transition from the current fuel cycle, and evaluate performance of nuclear systems and facilities in place during the transition. These studies also help identify any barriers to achieve transition. Oak Ridge National Laboratory (ORNL) Fuel Cycle Options Campaign team used ORION to analyze the transition pathway from the existing US nuclear fuel cycle—the once-through use of low-enriched-uranium (LEU) fuel in thermal-spectrum light water reactors (LWRs) —to a new fuel cycle with continuous recycling of plutonium and uranium in sodium fast reactors (SFRs). This paper discusses the analysis of the transition from an LWR to an SFR fleet using ORION, highlights the role of lifetime extensions of existing LWRs to aid transition, and discusses how a slight delay in SFR deployment can actually reduce the time to achieve an equilibrium fuel cycle.

  19. ENVIRONMENTAL ASSESSMENT METHODOLOGY FOR THE NUCLEAR FUEL CYCLE

    Energy Technology Data Exchange (ETDEWEB)

    Brenchley, D. L.; Soldat, J. K.; McNeese, J. A.; Watson, E. C.

    1977-07-01

    This report describes the methodology for determining where environmental control technology is required for the nuclear fuel cycle. The methodology addresses routine emission of chemical and radioactive effluents, and applies to mining, milling, conversion, enrichment, fuel fabrication, reactors (LWR and BWR) and fuel reprocessing. Chemical and radioactive effluents are evaluated independently. Radioactive effluents are evaluated on the basis of maximum exposed individual dose and population dose calculations for a 1-year emission period and a 50-year commitment. Sources of radionuclides for each facility are then listed according to their relative contribution to the total calculated dose. Effluent, ambient and toxicology standards are used to evaluate the effect of chemical effluents. First, each chemical and source configuration is determined. Sources are tagged if they exceed existirrg standards. The combined effect of all chemicals is assessed for each facility. If the additive effects are unacceptable, then additional control technology is recommended. Finally, sources and their chemicals at each facility are ranked according to their relative contribution to the ambient pollution level. This ranking identifies those sources most in need of environmental control.

  20. Coated Particle and Deep Burn Fuels Monthly Highlights December 2010

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Lance Lewis [ORNL; Bell, Gary L [ORNL; Besmann, Theodore M [ORNL

    2011-01-01

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

  1. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and nucl......A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...... and nuclear fuel-based energy technologies....

  2. Magnetic Biocomposites for Remote Melting.

    Science.gov (United States)

    Zhou, Mengbo; Liebert, Tim; Müller, Robert; Dellith, Andrea; Gräfe, Christine; Clement, Joachim H; Heinze, Thomas

    2015-08-10

    A new approach toward the fabrication of biocompatible composites suitable for remote melting is presented. It is shown that magnetite nanoparticles (MNP) can be embedded into a matrix of biocompatible thermoplastic dextran esters. For that purpose, fatty acid esters of dextran with adjustable melting points in the range of 30-140 °C were synthesized. Esterification of the polysaccharide by activation of the acid as iminium chlorides guaranteed mild reaction conditions leading to high quality products as confirmed by FTIR- and NMR spectroscopy as well as by gel permeation chromatography (GPC). A method for the preparation of magnetically responsive bionanocomposites was developed consisting of combined dissolution/suspension of the dextran ester and hydrophobized MNPs in an organic solvent followed by homogenization with ultrasonication, casting of the solution, drying and melting of the composite for a defined shaping. This process leads to a uniform distribution of MNPs in nanocomposite as revealed by scanning electron microscope. Samples of different geometries were exposed to high frequency alternating magnetic field. It could be shown that defined remote melting of such biocompatible nanocomposites is possible for the first time. This may lead to a new class of magnetic remote control systems, which are suitable for controlled release applications or self-healing materials.

  3. UNCONSTRAINED MELTING AND SOLIDIFICATION INSIDE ...

    African Journals Online (AJOL)

    2015-09-01

    Sep 1, 2015 ... Department of Mechanical Engineering, Sari Branch, Islamic Azad University, ... at initial time of melting process where the layer of liquid PCM near hot ... They carried out the simulation at different Rayleigh numbers ranging from 10 .... An enthalpy-porosity technique [28] is used in FLUENT for modeling the.

  4. Thermodynamics of freezing and melting

    DEFF Research Database (Denmark)

    Pedersen, Ulf Rørbæk; Costigliola, Lorenzo; Bailey, Nicholas

    2016-01-01

    phases at a single thermodynamic state point provide the basis for calculating the pressure, density and entropy of fusion as functions of temperature along the melting line, as well as the variation along this line of the reduced crystalline vibrational mean-square displacement (the Lindemann ratio...

  5. A novel design approach for a neutron measurement station for burnt fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dietler, Rodolfo, E-mail: rodolfo.dietler@axpo.ch [Axpo AG Kernenergie, CH-5401 Baden (Switzerland); Hursin, Mathieu, E-mail: mathieu.hursin@psi.ch [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Perret, Gregory, E-mail: gregory.perret@psi.ch [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Jordan, Kelly, E-mail: kjordan@mse.ufl.edu [University of Florida, 180 Rhines Hall, PO Box 116400, Gainesville, FL 32611-6400 (United States); Chawla, Rakesh, E-mail: rakesh.chawla@epfl.ch [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2012-11-21

    The design and characterization of a passive neutron measurement station for highly burnt fuel has been undertaken at the Paul Scherrer Institute (PSI). The measurement station aims at the determination of the total neutron emission rate of full-length light water reactor (LWR) fuel rods, as also the corresponding axial distributions. It is intended that the measurement station be introduced into the hot cells available at PSI to allow measuring the neutron emission of spent fuel rods provided by the Swiss nuclear power plants. In addition, the neutron emission of a large set of burnt fuel samples that have been previously characterized by post-irradiation examination (PIE) will be measured, in order to relate neutron emission to the burnup and isotopic composition of different fuel types. The design of the measurement station is presented in this article. A post-processing algorithm is introduced to improve the spatial resolution of the 'measured' axial profile. In order to quantify the accuracy of the reconstructed neutron source distribution, a figure-of-merit (FOM) is defined and adapted to the detection procedure. With the optimized measurement station and procedure, it is estimated that the neutron emission distribution of a highly burnt, full-length fuel rod would be measurable with acceptable accuracy in about 20 min.

  6. The generation of denatured reactor plutonium by different options of the fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Broeders, C.H.M.; Kessler, G. [Inst. for Neutron Physics and Reactor Technology, Research Center Karlsruhe (Germany)

    2006-11-15

    Denatured (proliferation resistant) reactor plutonium can be generated in a number of different fuel cycle options. First denatured reactor plutonium can be obtained if, instead of low enriched U-235 PWR fuel, re-enriched U-235/U-236 from reprocessed uranium is used (fuel type A). Also the envisaged existing 2,500 t of reactor plutonium (being generated world wide up to the year 2010), mostly stored in intermediate fuel storage facilities at present, could be converted during a transition phase into denatured reactor plutonium by the options fuel type B and D. Denatured reactor plutonium could have the same safeguards standard as present low enriched (<20% U-235) LWR fuel. It could be incinerated by recycling once or twice in PWRs and subsequently by multi-recycling in FRs (CAPRA type or IFRs). Once denatured, such reactor plutonium could remain denatured during multiple recycling. In a PWR, e.g., denatured reactor plutonium could be destroyed at a rate of about 250 kg/GWey. While denatured reactor plutonium could be recycled and incinerated under relieved IAEA safeguards, neptunium would still have to be monitored by the IAEA in future for all cases in which considerable amounts of neptunium are produced. (orig.)

  7. A novel design approach for a neutron measurement station for burnt fuel

    Science.gov (United States)

    Dietler, Rodolfo; Hursin, Mathieu; Perret, Gregory; Jordan, Kelly; Chawla, Rakesh

    2012-11-01

    The design and characterization of a passive neutron measurement station for highly burnt fuel has been undertaken at the Paul Scherrer Institute (PSI). The measurement station aims at the determination of the total neutron emission rate of full-length light water reactor (LWR) fuel rods, as also the corresponding axial distributions. It is intended that the measurement station be introduced into the hot cells available at PSI to allow measuring the neutron emission of spent fuel rods provided by the Swiss nuclear power plants. In addition, the neutron emission of a large set of burnt fuel samples that have been previously characterized by post-irradiation examination (PIE) will be measured, in order to relate neutron emission to the burnup and isotopic composition of different fuel types. The design of the measurement station is presented in this article. A post-processing algorithm is introduced to improve the spatial resolution of the "measured" axial profile. In order to quantify the accuracy of the reconstructed neutron source distribution, a figure-of-merit (FOM) is defined and adapted to the detection procedure. With the optimized measurement station and procedure, it is estimated that the neutron emission distribution of a highly burnt, full-length fuel rod would be measurable with acceptable accuracy in about 20 min.

  8. The Design Features of the Double-Banked AMBIDEXTER Utilizing DUPIC Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Tae Kyu [KHNP Central Research Institute, Daejeon (Korea, Republic of); Lee, Young Joon; Hong, Sung Taek [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Seo, Myung Hwan [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of); Kwon, Tae An [KHNP, Daejeon (Korea, Republic of); Oh, Se Kee [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-10-15

    Since the on-site spent fuel storage capabilities at reactors in Korea are expected to be saturated in a few years, the government has been pressed to find a solution for the spent nuclear fuel. So far one of workable means for reducing the load would be utilizing DUPIC fuel cycle technology. The technology was developed through Korea-Canada-U.S. collaboration to utilize the LWR spent fuel for the CANDU reactor. However, by various sociopolitical reasons, the DUPIC technology has not been yet commercialized. As the other alternatives to use the DUPIC technology, Gen-IV reactors would be pertinent. In the following session, the design features of a molten salt reactor system that can burn DUPIC fuel are explained. The followings are derived as conclusions after considering all the factors; The AMDEC, compared to ORIGEN2 simulations, can calculate the nuclides concentration changes within 1% deviation in various core zones and reactor system components by using different library sets which are weighted with each neutron spectrum; Fuel-flow effects coupled with nuclear reactions is well reflected in the AMDEC.

  9. Metrics for the Evaluation of Light Water Reactor Accident Tolerant Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton

    2001-09-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of LWRs became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of accident tolerant fuel (ATF) development is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. The U.S. Department of Energy is sponsoring multiple teams to develop ATF concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This paper summarizes technical evaluation methodology proposed in the U.S. to aid in the optimization and down-selection of candidate ATF designs. This methodology will continue to be refined via input from the research community and industry, such that it is available to support the planned down-selection of ATF concepts in 2016.

  10. Compact Fuel Element Environment Test

    Science.gov (United States)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

    2012-01-01

    Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

  11. Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Galloway, Jack D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-12

    In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermal swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.

  12. High-Efficiency, High-Capacity, Low-NOx Aluminum Melting Using Oxygen-Enhanced Combustion

    Energy Technology Data Exchange (ETDEWEB)

    D' Agostini, M.D.

    2000-06-02

    This report describes the development and application of a novel oxygen enhanced combustion system with an integrated vacuum swing adsorption (VSA) oxygen supply providing efficient, low NOx melting in secondary aluminum furnaces. The mainstay of the combustion system is a novel air-oxy-natural gas burner that achieves high productivity and energy efficiency with low NOx emissions through advanced mixing concepts and the use of separate high- and low-purity oxidizer streams. The technology was installed on a reverberatory, secondary aluminum melting plant at the Wabash Aluminum Alloy's Syracuse, N.Y. plant, where it is currently in operation. Field testing gave evidence that the new burner technology meets the stringent NOx emissions target of 0.323 lb NO2/ton aluminum, thus complying with regulations promulgated by Southern California's South Coast Air Quality Management District (SCAQMD). Test results also indicated that the burner technology exceeded fuel efficiency and melting capacity goals. Economic modeling showed that the novel air-oxy-fuel (ADF) combustion technology provides a substantial increase in furnace profitability relative to air-fuel operation. Model results also suggest favorable economics for the air-oxy-fuel technology relative to a full oxy-fuel conversion of the furnace.

  13. Results of recent KROTOS FCI tests. Alumina vs. corium melts

    Energy Technology Data Exchange (ETDEWEB)

    Huhtiniemi, I.; Magallon, D.; Hohmann, H. [Commission of the European Communities, Ispra (Italy). Joint Research Center

    1998-01-01

    Recent results from KROTOS fuel-coolant interaction experiments are discussed. Five tests with alumina were performed under highly subcooled conditions, all of these tests resulted in spontaneous steam explosions. Additionally, four tests were performed at low subcooling to confirm, on one hand, the suppression of spontaneous steam explosions under such conditions and, on the other hand, that such a system is still triggerable using an external initiator. The other test parameters in these alumina tests included the melt superheat and the initial pressure. All the tests in the investigated superheat range (150 K - 750 K) produced a steam explosion and no evidence of the explosion suppression by the elevated initial pressure (in the limited range of 0.1 - 0.375 MPa) was observed in the alumina tests. The corium test series include a test with 3 kg of melt under both subcooled and near saturated conditions at ambient pressure. Two additional tests were performed with subcooled water; one test was performed at an elevated pressure of 0.2 MPa with 2.4 kg of melt and another test with 5.1 kg of melt at ambient pressure. None of these tests with corium produced a propagating energetic steam explosion. However, propagating low energy (about twice the energy of the trigger pulse) events were observed. All corium tests produced significantly higher water level swells during the mixing phase than the corresponding alumina tests. Present experimental evidence suggests that the water depletion in the mixing zone suppresses energetic steam explosions with corium melts at ambient pressure and in the present pour geometry. Processes that could produce such a difference in void generation are discussed. (author)

  14. Fuel cells

    Directory of Open Access Journals (Sweden)

    D. N. Srivastava

    1962-05-01

    Full Text Available The current state of development of fuel cells as potential power sources is reviewed. Applications in special fields with particular reference to military requirements are pointed out.

  15. Future Fuels

    Science.gov (United States)

    2006-04-01

    Storage Devices, Fuel Management, Gasification, Fischer-Tropsch, Syngas , Hubberts’s Peak UNCLAS UNCLAS UNCLAS UU 80 Dr. Sujata Millick (703) 696...prices ever higher, and perhaps lead to intermittent fuel shortages as production fluctuates. Clearly, this competition for resources also provides oil...producers multiple options for selling their products, and raises the possibility that the US could face shortages resulting from shifts in

  16. A study of fuel behavior under reactivity initiated accident conditions — review

    Science.gov (United States)

    Ishikawa, Michio; Shiozawa, Shusaku

    1980-11-01

    Results obtained in the 400 tests performed to simulate reactivity initiated accidents since 1975 in the Japanese Nuclear Safety Research Reactor, are described. Tests included the effects of cooling environment, defective fuel elements, fuel design parameters, the behaviour of fuel elements for various reactor types, all done for a wide range of energy deposition. Four types of basic fuel failure mechanisms have been established, and are discussed in detail: cladding melt failure, UO 2 melt failure, high temperature burst failure and low temperature burst failure. Future test plans up to 1990 are out-lined and features requiring particular attention are pointed out.

  17. Basal terraces on melting ice shelves

    National Research Council Canada - National Science Library

    Dutrieux, Pierre; Stewart, Craig; Jenkins, Adrian; Nicholls, Keith W; Corr, Hugh F. J; Rignot, Eric; Steffen, Konrad

    2014-01-01

    Ocean waters melt the margins of Antarctic and Greenland glaciers, and individual glaciers' responses and the integrity of their ice shelves are expected to depend on the spatial distribution of melt...

  18. Filling Knowledge Gaps with Five Fuel Cycle Studies

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Jess Gehin; William Halsey; Temitope Taiwo

    2010-11-01

    the five studies was spotty. Some studies did not have existing or past work to draw on in one or more of these areas. Resource constraints limited the amount of new analyses that could be performed. Little or no assessment was done of how soon any of the technologies could be deployed and therefore how quickly they could impact domestic or international fuel cycle performance. There were six common R&D needs, such as the value of advanced fuels, cladding, coating, and structure that would survive high neutron fluence. When a technology family is considered for use in a new fuel cycle mission, fuel cycle performance characteristics are dependent on both the design choices and the fuel cycle approach. For example, the use of the sodium-cooled fast reactor to provide recycle in either breeder or burner mode has been studied for decades, but the SFR could be considered for once-through fuel cycle with the physical reactor design and fuel management parameters changed. In addition, the sustained recycle with Th/U-233 in LWR could be achieved with a heterogeneous assembly and derated power density. Therefore, it may or may not be adjustable for other fuel cycle missions although a reactor intended for one fuel cycle mission is built. Simple parameter adjustment in applying a technology family to a new fuel cycle mission should be avoided and, if observed, the results viewed with caution.

  19. An innovative energy-saving in-flight melting technology and its application to glass production

    Directory of Open Access Journals (Sweden)

    Yaochun Yao et al

    2008-01-01

    Full Text Available The conventional method used for glass melting is air-fuel firing, which is inefficient, energy-intensive and time-consuming. In this study, an innovative in-flight melting technology was developed and applied to glass production for the purposes of energy conservation and environmental protection. Three types of heating sources, radio-frequency (RF plasma, a 12-phase alternating current (ac arc and an oxygen burner, were used to investigate the in-flight melting behavior of granulated powders. Results show that the melted particles are spherical with a smooth surface and compact structure. The diameter of the melted particles is about 50% of that of the original powders. The decomposition and vitrification degrees of the prepared powders decrease in the order of powders prepared by RF plasma, the 12-phase ac arc and the oxygen burner. The largest heat transfer is from RF plasma to particles, which results in the highest particle temperature (1810 °C and the greatest vitrification degree of the raw material. The high decomposition and vitrification degrees, which are achieved in milliseconds, shorten the melting and fining times of the glass considerably. Our results indicate that the proposed in-flight melting technology is a promising method for use in the glass industry.

  20. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Chi Thanh

    2009-09-15

    Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents. In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment. The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis. The CFD method, on the one hand

  1. AECL/US INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors -- Fuel Requirements and Down-Select Report

    Energy Technology Data Exchange (ETDEWEB)

    William Carmack; Randy D. Lee; Pavel Medvedev; Mitch Meyer; Michael Todosow; Holly B. Hamilton; Juan Nino; Simon Philpot; James Tulenko

    2005-06-01

    The U.S. Advanced Fuel Cycle Program and the Atomic Energy Canada Ltd (AECL) seek to develop and demonstrate the technologies needed to minimize the overall Pu and minor actinides present in the light water reactor (LWR) nuclear fuel cycles. It is proposed to reuse the Pu from LWR spent fuel both for the energy it contains and to decrease the hazard and proliferation impact resulting from storage of the Pu and minor actinides. The use of fuel compositions with a combination of U and Pu oxide (MOX) has been proposed as a way to recycle Pu and/or minor actinides in LWRs. It has also been proposed to replace the fertile U{sup 238} matrix of MOX with a fertile-free matrix (IMF) to reduce the production of Pu{sup 239} in the fuel system. It is important to demonstrate the performance of these fuels with the appropriate mixture of isotopes and determine what impact there might be from trace elements or contaminants. Previous work has already been done to look at weapons-grade (WG) Pu in the MOX configuration [1][2] and the reactor-grade (RG) Pu in a MOX configuration including small (4000 ppm additions of Neptunium). This program will add to the existing database by developing a wide variety of MOX fuel compositions along with new fuel compositions called inert-matrix fuel (IMF). The goal of this program is to determine the general fabrication and irradiation behavior of the proposed IMF fuel compositions. Successful performance of these compositions will lead to further selection and development of IMF for use in LWRs. This experiment will also test various inert matrix material compositions with and without quantities of the minor actinides Americium and Neptunium to determine feasibility of incorporation into the fuel matrices for destruction. There is interest in the U.S. and world-wide in the investigation of IMF (inert matrix fuels) for scenarios involving stabilization or burn down of plutonium in the fleet of existing commercial power reactors. IMF offer the

  2. Filament stretching rheometry of polymer melts

    DEFF Research Database (Denmark)

    Hassager, Ole; Nielsen, Jens Kromann; Rasmussen, Henrik Koblitz

    2005-01-01

    The Filament Stretching Rheometry (FSR) method developed by Sridhar, McKinley and coworkers for polymer solutions has been extended to be used also for polymer melts. The design of a melt-FSR will be described and differences to conventional melt elongational rheometers will be pointed out. Results...

  3. Filament stretching rheometry of polymer melts

    DEFF Research Database (Denmark)

    Hassager, Ole; Nielsen, Jens Kromann; Rasmussen, Henrik Koblitz

    2005-01-01

    The Filament Stretching Rheometry (FSR) method developed by Sridhar, McKinley and coworkers for polymer solutions has been extended to be used also for polymer melts. The design of a melt-FSR will be described and differences to conventional melt elongational rheometers will be pointed out. Results...

  4. SURFACE MELTING OF ALUMINIUM ALLOYS

    OpenAIRE

    Veit, S.; Albert, D; Mergen, R.

    1987-01-01

    The wear properties of aluminium base alloys are relatively poor. Laser surface melting and alloying has proved successful in many alloy systems as a means of significantly improving the surface properties. The present work describes experiments designed to establish the scope of laser treatment of aluminium alloys. Aluminium does not absorb CO2 laser light as well as other metals which necessitated first a general study of absorption caotings. Aluminium alloys offer fewer opportunities than ...

  5. Melting Behaviour of Ferronickel Slags

    Science.gov (United States)

    Sagadin, Christoph; Luidold, Stefan; Wagner, Christoph; Wenzl, Christine

    2016-12-01

    The industrial manufacturing of ferronickel in electric furnaces produces large amounts of slag with strong acidic character and high melting points, which seriously stresses the furnace refractory lining. In this study, the melting behavior of synthetically produced ferronickel slags on magnesia as refractory material was determined by means of a hot stage microscope. Therefore, slags comprising the main oxides SiO2 (35-70 wt.%), MgO (15-45 wt.%) and Fe2O3 (5-35 wt.%) were melted in a graphite crucible and afterwards analyzed by a hot stage microscope. The design of experiments, which was created by the statistic software MODDE®, included 20 experiments with varying slag compositions as well as atmospheres. The evaluation of the test results occurred at three different characteristic states of the samples like the softening point according to DIN 51730 and the temperatures at which the area of residual cross-section of the samples amounted to 30% and 40%, respectively, of the original value depending of their SiO2/MgO ratio and iron oxide content. Additionally, the thickness of the zone influenced by the slag was measured and evaluated.

  6. Optimization of the Brass Melting

    Directory of Open Access Journals (Sweden)

    Biernat S.

    2014-08-01

    Full Text Available The article describes the optimization of the melting brass. Brasses, as one of the most popular alloys of copper, deserve special attention in the context of the processes of melting, which in turn would provide not only products of better quality, but also reduce the cost of their production or refining. For this purpose, several studies carried out deriatographic (DTA and thermogravimetric (TG using derivatograph. The results were confronted with the program SLAG - PROP used to evaluate the physicochemical properties of the coatings extraction. Based on the survey and analysis of the program can identify the most favorable physico - chemical properties, which should be carried out treatments. This allows for slag mixtures referred configurations oxide matrix containing specific stimulators of the reaction. Conducted empirical studies indicate a convergence of the areas proposed by the application. It should also be noted that the program also indicates additional areas in which to carry out these processes would get even better, to optimize the melting process, the results.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  8. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

  9. Hybrid rocket fuel combustion and regression rate study

    Science.gov (United States)

    Strand, L. D.; Ray, R. L.; Anderson, F. A.; Cohen, N. S.

    1992-01-01

    The objectives of this study are to develop hybrid fuels (1) with higher regression rates and reduced dependence on fuel grain geometry and (2) that maximize potential specific impulse using low-cost materials. A hybrid slab window motor system was developed to screen candidate fuels - their combustion behavior and regression rate. Combustion behavior diagnostics consisted of video and high speed motion pictures coverage. The mean fuel regression rates were determined by before and after measurements of the fuel slabs. The fuel for this initial investigation consisted of hydroxyl-terminated polybutadiene binder with coal and aluminum fillers. At low oxidizer flux levels (and corresponding fuel regression rates) the filled-binder fuels burn in a layered fashion, forming an aluminum containing binder/coal surface melt that, in turn, forms into filigrees or flakes that are stripped off by the crossflow. This melt process appears to diminish with increasing oxidizer flux level. Heat transfer by radiation is a significant contributor, producing the desired increase in magnitude and reduction in flow dependency (power law exponent) of the fuel regression rate.

  10. Solar fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bolton, J.R.

    1978-11-17

    The paper is concerned with (1) the thermodynamic and kinetic limits for the photochemical conversion and storage of solar energy as it is received on the earth's surface, and (2) the evaluation of a number of possible photochemical reactions with particular emphasis on the production of solar hydrogen from water. Procedures for generating hydrogen fuel are considered. Topics examined include the general requirements for a fuel-generation reaction, the photochemical reaction, limits on the conversion of light energy to chemical energy, an estimate of chemical storage efficiency, and the water decomposition reaction.

  11. Neutronic Analysis and Transmutation Performance of Th-based Molten Salt Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bak, Sangin; Chai, Jongseo; Hong, Seungwoo; Kadi, Yacine; Tenreiro, Claudio [Department of Energy Science/SungKyunKwan Univ., Suwon (Korea, Republic of); Curbelo, Jesus Perez [Higher Institute of Technologies and Applied Sciences, Havana (Chad); Kadi, Yacine [Engineering Department/European Organization for Nuclear Research, Geneva (Swaziland); Tenreiro, Claudio [Faculty of Engineering/Univ. of Talca, Talca (Chile)

    2013-05-15

    The utilization of thorium in various reactors ranging from thermal to fast neutron energies has been successfully demonstrated, but substantial R and D is required before commercialization becomes possible. The molten-salt reactor (MSR) systems present the very special feature of a liquid fuel. MSR concepts, which can be used as efficient burners of TRU from spent LWR fuel, have also a breeding capability in any kind of neutron spectrum (from thermal to fast), when using the thorium fuel cycle. It has a very interesting potential for the minimization of radio toxic nuclear waste. For a pure Th-U fuel, thermal/epithermal molten salt is better than a fast one. The system has better breeding ratio and can reach k{sub eff} with smaller enrichment than fast system, because of the neutron leakage of the fast system. However the fast fuel salts is preferred for the transmutation of Minor Actinides. When mixed with MA and Pu, thorium based fast molten salts have similar breeding capability as thermal/epithermal salts.

  12. Verification and Validation of the BISON Fuel Performance Code for PCMI Applications

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle Allan Lawrence [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Gardner, Russell James [Idaho National Laboratory; Perez, Danielle Marie [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-06-01

    BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described. Validation for application to light water reactor (LWR) PCMI problems is assessed by comparing predicted and measured rod diameter following base irradiation and power ramps. Results indicate a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

  13. Inherent safety advantages of carbide fuel systems and technical issues regarding natural convection in LMRs

    Energy Technology Data Exchange (ETDEWEB)

    Barthold, W.P. [Barthold and Associates, Inc., Albuquerque, NM (United States)

    1984-08-01

    The scope of work is to summarize inherent safety advantages that are unique to the use of a carbide based fuel system and to summarize the technical issues regarding natural convection flow in LMFBR cores. As discussed in this report, carbide fuel provides the designer with far greater flexibility than oxide fuel. Carbide fuel systems can be designed to eliminate major accident initiators. They turn quantitative advantages into a qualitative advantage. The author proposed to LANL a series of core design and component concepts that would greatly enhance the safety of carbide over oxide systems. This report cites a series of safety advantages which potentially exist for a carbide fuel system. Natural convection issues have not been given much attention in the past. Only during the last few years has this issue been addressed in some detail. Despite claims to the contrary by some of the LMR contractors, the author does not think that the natural convection phenomena is fully understood. Some of the approximations made in natural convection transient analyses have probably a greater impact on calculated transient temperatures than the effects under investigation. Only integral in-pile experimental data and single assembly out-of-pile detailed data are available for comparisons with analytical models and correlations. Especially for derated cores, the natural convection capability of a LMR should be far superior to that of a LWR. The author ranks the natural convection capability of the LMR as the most important inherent safety feature.

  14. Neutronic Analyses for HEU to LEU fuel conversion of the Massachusetts Institute of Technology.

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, E. H.; Newton, T. H.; Bergeron, A.; Horelik, N.; Stevens, J. G (Nuclear Engineering Division); ( NS)

    2011-03-02

    The Massachusetts Institute of Technology (MIT) reactor (MITR-II), based in Cambridge, Massachusetts, is a research reactor designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the MITR-II. This report presents the results of steady state neutronic safety analyses for conversion of MITR-II from the use of HEU fuel to the use of U-Mo LEU fuel. The objective of this work was to demonstrate that the safety analyses meet current requirements for an LEU core replacement of MITR-II.

  15. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  16. Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Fariz Abdul; Lee, John C. [University of Michigan, Ann Arbor, MI (United States); Franceschini, Fausto; Wenner, Michael [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2012-07-01

    As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning the legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and

  17. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Čerba, Štefan, E-mail: stefan.cerba@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Vrban, Branislav; Lüley, Jakub [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Dařílek, Petr [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Zajac, Radoslav, E-mail: radoslav.zajac@vuje.sk [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Nečas, Vladimír; Haščik, Ján [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia)

    2014-02-15

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.

  18. An Advanced Option for Sodium Cooled TRU Burner Loaded with Uranium-Free Fuels

    Energy Technology Data Exchange (ETDEWEB)

    You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    The sodium cooled fast reactors of this kind that are called burners are designed to have low conversion ratio by reducing fuel volume fraction or reducing neutron leakage or increasing neutron absorption. However, the typical SFR burners have a limited ability of TRU burning rate due to the fact that they use metallic or oxide fuels containing fertile nuclides such as {sup 238}U and {sup 232}Th and these fertile nuclides generate fissile nuclides through neutron capture even if they are designed to have low conversion ratio (e.g., 0.6). To further enhance the TRU burning rate, the removal of the fertile nuclides from the initial fuels is required and it will accelerate the reduction of TRUs that are accumulated in storages of LWR spent fuels. However, it has been well-known 4 that the removals of the fertile nuclides from the fuel degrade the inherent safety of the SFR burner cores through the significant decrease of the fuel Doppler effect, the increase of sodium void reactivity worth, and reduction of delayed neutron fraction. In this work, new option for the sodium cooled fast TRU burner cores loaded with fertile-free metallic fuels was proposed and the new cores were designed by using the suggested option. The cores were designed to enhance the inherent safety characteristics by using axially central absorber region and 6 or 12 ZrH1.8 moderator rods per fuel assembly. For each option, we considered two different types of fertile-free ternary metallic fuel (i.e., TRU-W-10Zr and TRU-Ni-10Zr). Also, we performed the BOR (Balance of Reactivity) analyses to show the self-controllability under ATWS as a measure of inherent safety. The core performance analysis showed that the new cores using axially central absorber region substantially improve the core performance parameters such as burnup reactivity swing and sodium void reactivity worth.

  19. Perspectives on the closed fuel cycle Implications for high-level waste matrices

    Science.gov (United States)

    Gras, Jean-Marie; Quang, Richard Do; Masson, Hervé; Lieven, Thierry; Ferry, Cécile; Poinssot, Christophe; Debes, Michel; Delbecq, Jean-Michel

    2007-05-01

    Nuclear energy accounts for 80% of electricity production in France, generating approximately 1150 t of spent fuel for an electrical output of 420 TWh. Based on a reprocessing-conditioning-recycling strategy, the orientations taken by Électricité de France (EDF) for the mid-term and the far-future are to keep the fleet performances at the highest level, and to maintain the nuclear option fully open by the replacement of present pressurized water reactor (PWR) by new light water reactor (LWR), such as the evolutionary pressurized reactor (EPR) and future Generation IV designs. Adaptations of waste materials to new requirements will come with these orientations in order to meet long-term energy sustainability. In particular, waste materials and spent fuels are expected to meet increased requirements in comparison with the present situation. So the treatment of higher burn-up UO2 spent fuel and MOX fuel requires determining the performances of glass and other matrices according to several criteria: chemical 'digestibility' (i.e. capacity of glass to incorporate fission products and minor actinides without loss of quality), resistance to alpha self-irradiation, residual power in view of disposal. Considering the long-term evolution of spent MOX fuel in storage, the helium production, the influence of irradiation damages accumulation and the evolution of the microstructure of the fuel pellet need to be known, as well as for the future fuels. Further, the eventual transmutation of minor actinides in fast neutron reactors (FR) of Generation IV, if its interest in optimising high-level waste management is proven, may also raise new challenges about the materials and fuel design. Some major questions in terms of waste materials and spent fuel are discussed in this paper.

  20. Vanadate-Sulfate Melt Thermochemistry Relating to Hot Corrosion of Thermal Barrier Coatings.

    Science.gov (United States)

    2007-11-02

    ship propulsion or electricity generation), such TBCs may be corroded by molten vanadate-sulfate deposits from fuel impurities. This Report provides a synopsis of vanadate-sulfate thermochemistry relating to TBC hot corrosion, and summarizes research conducted on this topic at the Naval Research Laboratory. The interactions of Na2O, V2O5 and SO3, the melt components which determine the composition of the vanadate-sulfate deposits, were examined and clarified. Vanadate-sulfate melts were shown to be nonideal rather than ideal a point of some contention in the past

  1. Fuel Cells

    Science.gov (United States)

    Hawkins, M. D.

    1973-01-01

    Discusses the theories, construction, operation, types, and advantages of fuel cells developed by the American space programs. Indicates that the cell is an ideal small-scale power source characterized by its compactness, high efficiency, reliability, and freedom from polluting fumes. (CC)

  2. Transport fuel

    DEFF Research Database (Denmark)

    Ronsse, Frederik; Jørgensen, Henning; Schüßler, Ingmar

    2014-01-01

    Worldwide, the use of transport fuel derived from biomass increased four-fold between 2003 and 2012. Mainly based on food resources, these conventional biofuels did not achieve the expected emission savings and contributed to higher prices for food commod - ities, especially maize and oilseeds...

  3. Rotation of melting ice disks due to melt fluid flow.

    Science.gov (United States)

    Dorbolo, S; Adami, N; Dubois, C; Caps, H; Vandewalle, N; Darbois-Texier, B

    2016-03-01

    We report experiments concerning the melting of ice disks (85 mm in diameter and 14 mm in height) at the surface of a thermalized water bath. During the melting, the ice disks undergo translational and rotational motions. In particular, the disks rotate. The rotation speed has been found to increase with the bath temperature. We investigated the flow under the bottom face of the ice disks by a particle image velocimetry technique. We find that the flow goes downwards and also rotates horizontally, so that a vertical vortex is generated under the ice disk. The proposed mechanism is the following. In the vicinity of the bottom face of the disk, the water eventually reaches the temperature of 4 °C for which the water density is maximum. The 4 °C water sinks and generates a downwards plume. The observed vertical vorticity results from the flow in the plume. Finally, by viscous entrainment, the horizontal rotation of the flow induces the solid rotation of the ice block. This mechanism seems generic: any vertical flow that generates a vortex will induce the rotation of a floating object.

  4. System Losses Study - FIT (Fuel-cycle Integration and Tradeoffs)

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Nick R. Soelberg; Samuel E. Bays; Robert S. Cherry; Denia Djokic; Candido Pereira; Layne F. Pincock; Eric L. Shaber; Melissa C. Teague; Gregory M. Teske; Kurt G. Vedros

    2010-09-01

    This team aimed to understand the broad implications of changes of operating performance and parameters of a fuel cycle component on the entire system. In particular, this report documents the study of the impact of changing the loss of fission products into recycled fuel and the loss of actinides into waste. When the effort started in spring 2009, an over-simplified statement of the objective was “the number of nines” – how would the cost of separation, fuel fabrication, and waste management change as the number of nines of separation efficiency changed. The intent was to determine the optimum “losses” of TRU into waste for the single system that had been the focus of the Global Nuclear Energy Program (GNEP), namely sustained recycle in burner fast reactors, fed by transuranic (TRU) material recovered from used LWR UOX-51 fuel. That objective proved to be neither possible (insufficient details or attention to the former GNEP options, change in national waste management strategy from a Yucca Mountain focus) nor appropriate given the 2009-2010 change to a science-based program considering a wider range of options. Indeed, the definition of “losses” itself changed from the loss of TRU into waste to a generic definition that a “loss” is any material that ends up where it is undesired. All streams from either separation or fuel fabrication are products; fuel feed streams must lead to fuels with tolerable impurities and waste streams must meet waste acceptance criteria (WAC) for one or more disposal sites. And, these losses are linked in the sense that as the loss of TRU into waste is reduced, often the loss or carryover of waste into TRU or uranium is increased. The effort has provided a mechanism for connecting these three Campaigns at a technical level that had not previously occurred – asking smarter and smarter questions, sometimes answering them, discussing assumptions, identifying R&D needs, and gaining new insights. The FIT model has been a

  5. Melt processing of radioactive waste: A technical overview

    Energy Technology Data Exchange (ETDEWEB)

    Schlienger, M.E.; Buckentin, J.M.; Damkroger, B.K.

    1997-04-01

    Nuclear operations have resulted in the accumulation of large quantities of contaminated metallic waste which are stored at various DOE, DOD, and commercial sites under the control of DOE and the Nuclear Regulatory Commission (NRC). This waste will accumulate at an increasing rate as commercial nuclear reactors built in the 1950s reach the end of their projected lives, as existing nuclear powered ships become obsolete or unneeded, and as various weapons plants and fuel processing facilities, such as the gaseous diffusion plants, are dismantled, repaired, or modernized. For example, recent estimates of available Radioactive Scrap Metal (RSM) in the DOE Nuclear Weapons Complex have suggested that as much as 700,000 tons of contaminated 304L stainless steel exist in the gaseous diffusion plants alone. Other high-value metals available in the DOE complex include copper, nickel, and zirconium. Melt processing for the decontamination of radioactive scrap metal has been the subject of much research. A major driving force for this research has been the possibility of reapplication of RSM, which is often very high-grade material containing large quantities of strategic elements. To date, several different single and multi-step melting processes have been proposed and evaluated for use as decontamination or recycling strategies. Each process offers a unique combination of strengths and weaknesses, and ultimately, no single melt processing scheme is optimum for all applications since processes must be evaluated based on the characteristics of the input feed stream and the desired output. This paper describes various melt decontamination processes and briefly reviews their application in developmental studies, full scale technical demonstrations, and industrial operations.

  6. Synthesis and sintering of UN-UO2 fuel composites

    Science.gov (United States)

    Jaques, Brian J.; Watkins, Jennifer; Croteau, Joseph R.; Alanko, Gordon A.; Tyburska-Püschel, Beata; Meyer, Mitch; Xu, Peng; Lahoda, Edward J.; Butt, Darryl P.

    2015-11-01

    The design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO2, which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO2 to UN has been suggested. In order to avoid carbon impurities, UN was synthesized from elemental uranium using a hydride-dehydride-nitride thermal synthesis route prior to mixing with up to 10 wt% UO2 in a planetary ball mill. UN and UN - UO2 composite pellets were sintered in Ar - (0-1 at%) N2 to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO2 composite pellets were also sintered in Ar - 100 ppm N2 to assess the effects of temperature (1700-2000 °C) on the final grain morphology and phase concentration.

  7. Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heidet, F.; Kim, T.; Grandy, C. (Nuclear Engineering Division)

    2012-07-30

    Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium is more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the

  8. Effect of melting conditions on striae in iron-bearing silicate melts

    DEFF Research Database (Denmark)

    Jensen, Martin; Yue, Yuanzheng

    2012-01-01

    Chemical striae are present in a broad range of glass products, but due to their negative impact on e.g., the optical and mechanical properties, elimination of striae from melts is a key issue in glass technology. By varying melting temperatures, retentions times and redox conditions of an iron......-bearing calciumaluminosilicate melt, we quantify the effect of each of the three melting parameters on the stria content in the melt. The quantification of the stria content in the melt is conducted by means of image analysis on casted melt samples. We find that in comparison to an extension of retention time an increase...... factors such as compositional fluctuation of melts and bubbling due to iron reduction on the stria content. During the melting process, striae with a chemical gradient in a more mobile species equilibrate faster than striae caused by a chemical gradient in a less mobile species. The temperature and time...

  9. Double-clad nuclear-fuel safety rod

    Science.gov (United States)

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  10. Fuel cycle analysis of once-through nuclear systems.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  11. Examination of nanosecond laser melting thresholds in refractory metals by shear wave acoustics

    Science.gov (United States)

    Abdullaev, A.; Muminov, B.; Rakhymzhanov, A.; Mynbayev, N.; Utegulov, Z. N.

    2017-07-01

    Nanosecond laser pulse-induced melting thresholds in refractory (Nb, Mo, Ta and W) metals are measured using detected laser-generated acoustic shear waves. Obtained melting threshold values were found to be scaled with corresponding melting point temperatures of investigated materials displaying dissimilar shearing behavior. The experiments were conducted with motorized control of the incident laser pulse energies with small and uniform energy increments to reach high measurement accuracy and real-time monitoring of the epicentral acoustic waveforms from the opposite side of irradiated sample plates. Measured results were found to be in good agreement with numerical finite element model solving coupled elastodynamic and thermal conduction governing equations on structured quadrilateral mesh. Solid-melt phase transition was handled by means of apparent heat capacity method. The onset of melting was attributed to vanished shear modulus and rapid radial molten pool propagation within laser-heated metal leading to preferential generation of transverse acoustic waves from sources surrounding the molten mass resulting in the delay of shear wave transit times. Developed laser-based technique aims for applications involving remote examination of rapid melting processes of materials present in harsh environment (e.g. spent nuclear fuels) with high spatio-temporal resolution.

  12. Metallic Fuel Casting Development and Parameter Optimization Simulations

    Energy Technology Data Exchange (ETDEWEB)

    R.S. Fielding; J. Crapps; C. Unal; J.R. Kennedy

    2013-03-01

    One of the advantages of metallic fuel is the abilility to cast the fuel slugs to near net shape with little additional processing. However, the high aspect ratio of the fuel is not ideal for casting. EBR-II fuel was cast using counter gravity injection casting (CGIC) but, concerns have been raised concerning the feasibility of this process for americium bearing alloys. The Fuel Cycle Research and Development program has begun developing gravity casting techniques suitable for fuel production. Compared to CGIC gravity casting does not require a large heel that then is recycled, does not require application of a vacuum during melting, and is conducive to re-usable molds. Development has included fabrication of two separate benchscale, approximately 300 grams, systems. To shorten development time computer simulations have been used to ensure mold and crucible designs are feasible and to identify which fluid properties most affect casting behavior and therefore require more characterization.

  13. Status of advanced carbide fuels: Past, present, and future

    Science.gov (United States)

    Anghaie, Samim; Knight, Travis

    2002-01-01

    Solid solution, mixed uranium/refractory metal carbide fuels such as (U, Zr, Nb)C, so called ternary carbide or tri-carbide fuels have great potential for applications in next generation advanced nuclear power reactors. Because of their high melting points, high thermal conductivity, improved resistance to hot hydrogen corrosion, and good fission product retention, these advanced nuclear fuels have great potential for high performance reactors with increased safety margins. Despite these many benefits, some concerns regarding carbide fuels include compatibility issues with coolant and/or cladding materials and their endurance under the extreme conditions associated with nuclear thermal propulsion. The status of these fuels is reviewed to characterize their performance for space nuclear power applications. Results of current investigations are presented and as well as future directions of study for these advanced nuclear fuels. .

  14. Fuel fragmentation model advances using TEXAS-V

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M.L.; El-Beshbeeshy, M.; Nilsuwankowsit, S.; Tang, J. [Wisconsin Univ., Madison, WI (United States). Dept. of Nuclear Engineering and Engineering Physics

    1998-01-01

    Because an energetic fuel-coolant interaction may be a safety hazard, experiments are being conducted to investigate the fuel-coolant mixing/quenching process (FARO) as well as the energetics of vapor explosion propagation for high temperature fuel melt simulants (KROTOS, WFCI, ZrEX). In both types of experiments, the dynamic breakup of the fuel is one of the key aspects that must be fundamentally understood to better estimate the magnitude of the mixing/quenching process or the explosion energetics. To aid our understanding the TEXAS fuel-coolant interaction computer model has been developed and is being used to analyze these experiments. Recently, the models for dynamic fuel fragmentation during the mixing and explosion phases of the FCI have been improved by further insights into these processes. The purpose of this paper is to describe these enhancements and to demonstrate their improvements by analysis of particular JRC FCI data. (author)

  15. Rheology of Melt-bearing Crustal Rocks

    Science.gov (United States)

    Rosenberg, C. L.; Medvedev, S.; Handy, M. R.

    2006-12-01

    A review and reinterpretation of previous experimental data on the deformation of melt-bearing crustal rocks (Rosenberg and Handy, 2005) revealed that the relationship of aggregate strength to melt fraction is non-linear, even if plotted on a linear ordinate and abscissa. At melt fractions, Φ 0.07, the dependence of aggregate strength on Φ is significantly greater than at Φ > 0.07. This melt fraction (Φ= 0.07) marks the transition from a significant increase in the proportion of melt-bearing grain boundaries up to this point to a minor increase thereafter. Therefore, we suggest that the increase of melt-interconnectivity causes the dramatic strength drop between the solidus and a melt fraction of 0.07. A second strength drop occurs at higher melt fractions and corresponds to the breakdown of the solid (crystal) framework, corresponding to the well-known "rheologically critical melt percentage" (RCMP; Arzi, 1978). Although the strength drop at the RCMP is about 4 orders of magnitude, the absolute value of this drop is small compared to the absolute strength of the unmelted aggregate, rendering the RCMP invisible in a linear aggregate strength vs. melt fraction diagram. Predicting the rheological properties and thresholds of melt-bearing crust on the basis of the results and interpretations above is very difficult, because the rheological data base was obtained from experiments performed at undrained conditions in the brittle field. These conditions are unlikely to represent the flow of partially melted crust. The measured strength of most of the experimentally deformed, partially-melted samples corresponds to their maximum differential stress, before the onset of brittle failure, not to their viscous strength during "ductile" (viscous) flow. To overcome these problems, we extrapolated a theoretically-derived flow law for partially melted granite deforming by diffusion-accommodated grain-boundary sliding (Paterson, 2001) and an experimentally-derived flow law for

  16. Viscosity model for aluminosilicate melt

    Directory of Open Access Journals (Sweden)

    Zhang G.H.

    2012-01-01

    Full Text Available The structurally based viscosity model proposed in our previous study is extended to include more components, e.g. SiO2, Al2O3, FeO, MnO, MgO, CaO, Na2O and K2O. A simple method is proposed to calculate the numbers of different types of oxygen ions classified by the different cations they bonded with, which is used to characterize the influence of composition on viscosity. When dealing with the aluminosilicate melts containing several basic oxides, the priority order is established for different cations for charge compensating Al3+ ions, according to the coulombic force between cation and oxygen anion. It is indicated that basic oxides have two paradox influences on viscosity: basic oxide with a higher basicity decreases viscosity more greatly by forming weaker non-bridging oxygen bond; while it increases viscosity more greatly by forming stronger bridging oxygen bond in tetrahedron after charge compensating Al3+ ion. The present model can extrapolate its application range to the system without SiO2. Furthermore, it could also give a satisfy interpretation to the abnormal phenomenon that viscosity increases when adding K2O to CaO-Al2O3-SiO2 melt within a certain composition range.

  17. Snow Melting and Freezing on Older Townhouses

    DEFF Research Database (Denmark)

    Nielsen, Anker; Claesson, Johan

    2011-01-01

    The snowy winter of 2009/2010 in Scandinavia prompted many newspaper articles on icicles falling from buildings and the risk this presented for people walking below. The problem starts with snow melting on the roof due to heat loss from the building. Melt water runs down the roof and some...... of it will freeze on the overhang. The rest of the water will either run off or freeze in gutters and downpipes or turn into icicles. This paper describes use of a model for the melting and freezing of snow on roofs. Important parameters are roof length, overhang length, heat resistance of roof and overhang......, outdoor and indoor temperature, snow thickness and thermal conductivity. If the snow thickness is above a specific limit value – the snow melting limit- some of the snow will melt. Another interesting limit value is the dripping limit. All the melt water will freeze on the overhang, if the snow thickness...

  18. RHEOLOGY FEATURE OF SIMPLE METAL MELT

    Institute of Scientific and Technical Information of China (English)

    C.J. Sun; H.R. Geng; Y.S. Shen; X.Y. Teng; Z.X. Yang

    2007-01-01

    The rheology feature of Sb, Bi melt and alloys was studied using coaxial cylinder high-temperature viscometer. The results showed that the curve of torsion-rotational speed for Sb melt presents a linear relation in all measured temperature ranges, whereas for the Bi melt, the curve presents obvious non-Newtonian feature within the low temperature range and at relative high shear stress. The rheology feature of Sb80Bi20 and Sb20Bi80, alloy melts was well correlated with that of Sb and Bi, respectively. It is considered that the rheology behavior of Sb melt plays a crucial role in Sb80Bi20, alloy and that of Bi melt plays a crucial role in Sb20Bi80 alloy.

  19. In-core measurements of fuel-clad interactions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett Peter

    2008-10-15

    A combination of on-line measurement techniques was used to demonstrate AOA in the Halden reactor: - Diameter gauge to demonstrate crud deposition - Coolant flow and temperature measurements to show effect of crud on thermal-hydraulic conditions - Neutron detectors to show power depression caused by boron in the crud. - Coolant chemistry analyses provided supporting evidence of AOA - Lithium return during shutdown - PIE showed that the type of crud observed in US plants suffering severe AOA can be reproduced. Loop systems allow testing under LWR thermal-hydraulic and water chemistry conditions. - A combination of on-line instrumentation allows measurements of complicated phenomena, egPWR AOA. - Techniques are under development to allow on-line measurements of fuel clad corrosion

  20. Development of Head-end Pyrochemical Reduction Process for Advanced Oxide Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Park, B. H.; Seo, C. S.; Hur, J. M.; Jeong, S. M.; Hong, S. S.; Choi, I. K.; Choung, W. M.; Kwon, K. C.; Lee, I. W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    The development of an electrolytic reduction technology for spent fuels in the form of oxide is of essence to introduce LWR SFs to a pyroprocessing. In this research, the technology was investigated to scale a reactor up, the electrochemical behaviors of FPs were studied to understand the process and a reaction rate data by using U{sub 3}O{sub 8} was obtained with a bench scale reactor. In a scale of 20 kgHM/batch reactor, U{sub 3}O{sub 8} and Simfuel were successfully reduced into metals. Electrochemical characteristics of LiBr, LiI and Li{sub 2}Se were measured in a bench scale reactor and an electrolytic reduction cell was modeled by a computational tool.

  1. An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, Cole A [ORNL; Godfrey, Andrew T [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Powers, Jeffrey J [ORNL; Maldonado, G Ivan [ORNL

    2014-01-01

    An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

  2. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

    1995-09-01

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  3. Multiscale Models of Melting Arctic Sea Ice

    Science.gov (United States)

    2014-09-30

    1 Multiscale Models of Melting Arctic Sea Ice Kenneth M. Golden University of Utah, Department of Mathematics phone: (801) 581-6851...feedback has played a major role in the recent declines of the summer Arctic sea ice pack. However, understanding the evolution of melt ponds and sea...Models of Melting Arctic Sea Ice 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) 5d. PROJECT NUMBER 5e. TASK NUMBER

  4. Fuel control system for dual fuel engines

    Energy Technology Data Exchange (ETDEWEB)

    Helmich, M.J.; Ryan, W.P.; Marvin, D.H.

    1987-11-24

    A fuel governing system for an engine adapted for operation on a first fuel and a second fuel is described comprising: a first fuel governing system including a spontaneous motion metering means; and a second fuel governing system, the second fuel governing system further comprising: means for providing a first signal indicative of position of the first fuel metering means, which signal approximates total load on the engine, means for providing a second signal of the selected percentage of first fuel relative to total load, means for controlling flow of the second fuel to the engine, which flow causes reflective displacement of the first fuel metering means, means for determining the difference between the first signal and the second signal, which difference is indicative of distance the first fuel metering means must be moved to attain the selected percentage of first fuel relative to total load, and means for causing operation of the means for controlling flow of the second fuel to the engine to cause displacement of the first fuel metering means equal to the distance the first fuel metering means must be moved to attain the selected percentage of first fuel relative to total load.

  5. Impact Melt in Small Lunar Highlands Craters

    Science.gov (United States)

    Plescia, J. B.; Cintala, M. J.; Robinson, M. S.; Barnouin, O.; Hawke, B. R.

    2011-01-01

    Impact-melt deposits are a typical characteristic of complex impact craters, occurring as thick pools on the crater floor, ponds on wall terraces, veneers on the walls, and flows outside and inside the rim. Studies of the distribution of impact melt suggested that such deposits are rare to absent in and around small (km to sub-km), simple impact craters. noted that the smallest lunar crater observed with impact melt was approximately 750 m in diameter. Similarly, theoretical models suggest that the amount of melt formed is a tiny fraction (crater volume and thus significant deposits would not be expected for small lunar craters. LRO LROC images show that impact-melt deposits can be recognized associated with many simple craters to diameters down to approximately 200 m. The melt forms pools on the crater floor, veneer on the crater walls or ejecta outside the crater. Such melt deposits are relatively rare, and can be recognized only in some fresh craters. These observations indicate that identifiable quantities of impact melt can be produced in small impacts and the presence of such deposits shows that the material can be aggregated into recognizable deposits. Further, the present of such melt indicates that small craters could be reliably radiometrically dated helping to constrain the recent impact flux.

  6. Low Melt Height Solidification of Superalloys

    Science.gov (United States)

    Montakhab, Mehdi; Bacak, Mert; Balikci, Ercan

    2016-06-01

    Effect of a reduced melt height in the directional solidification of a superalloy has been investigated by two methods: vertical Bridgman (VB) and vertical Bridgman with a submerged baffle (VBSB). The latter is a relatively new technique and provides a reduced melt height ahead of the solidifying interface. A low melt height leads to a larger primary dendrite arm spacing but a lower mushy length, melt-back transition length, and porosity. The VBSB technique yields up to 38 pct reduction in the porosity. This may improve a component's mechanical strength especially in a creep-fatigue type dynamic loading.

  7. Solute Redistribution in Directional Melting Process

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    @@The solute redistribution in directional melting process is theoretically studied. Based on quantitative evaluations, uniform solute distribution in liquid and a quasi-steady solute distribution in solid are supposed. The discussion on the solute balance comes to a simple model for the solute redistribution in directional melting process. As an example, the variation of liquid composition during melting process of carbon steel is quantitatively evaluated using the model. Results show that the melting of an alloy starts at solidus temperature, but approaches the liquidus temperature after a very short transient process.

  8. Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Wegst, Ulrike G.K. [Dartmouth College, Hanover, NH (United States). Thayer School of Engineering; Allen, Todd [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Univ. of Wisconsin, Madison, WI (United States)

    2014-04-07

    Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.

  9. Final report of fuel dynamics Test E7

    Energy Technology Data Exchange (ETDEWEB)

    Doerner, R.C.; Murphy, W.F.; Stanford, G.S.; Froehle, P.H.

    1977-04-01

    Test data from an in-pile failure experiment of high-power LMFBR-type fuel pins in a simulated $3/s transient-overpower (TOP) accident are reported and analyzed. Major conclusions are that (1) a series of cladding ruptures during the 100-ms period preceding fuel release injected small bursts of fission gas into the flow stream; (2) gas release influenced subsequent cladding melting and fuel release (there were no measurable FCI's (fuel-coolant interactions), and all fuel motion observed by the hodoscope was very slow); (3) the predominant postfailure fuel motion appears to be radial swelling that left a spongy fuel crust on the holder wall; (4) less than 4 to 6 percent of the fuel moved axially out of the original fuel zone, and most of this froze within a 10-cm region above the original top of the fuel zone to form the outlet blockage. An inlet blockage approximately 1 cm long was formed and consisted of large interconnected void regions. Both blockages began just beyond the ends of the fuel pellets.

  10. Preliminary analysis of modeling of Pars and steam injectors to support long-term operation of LWR passive ECCS using a best estimate thermal-hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Morales S, J. B.; Sanchez J, J. [UNAM, Facultad de Ingenieria, Circuito Interior s/n, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: jaimebmoraless@gmail.co [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2010-10-15

    In order to control the amount of hydrogen produced, during the chemical reaction of steam and strongly overheated zirconium in a LWR severe accident, several control methos have been proposed. In a considerable number of nuclear power plants the initial preferred solution was either the use of an inert containment atmosphere or the use of igniters. However, the use of Pars (Passive autocatalytic recombiners) has been considered an important addition to this set of H{sub 2} controls. Pars have appealing features such as, low maintenance, do not require active components to start operation, and can initiate the H{sub 2} and O{sub 2} recombination at low concentrations. However, under certain conditions these Pars may trigger H{sub 2} ignition leading to deflagration s with possible containment damages. Combustions, poisons and liquid may inhibit or reduce the Pars operation. Advanced LWR designs (generation III + and IV) are also considering the use of Pars as a supporting system to the long term operation of their passive ECCS (Emergency Core Cooling Systems) because they may transport the energy of containment H{sub 2} and O{sub 2} gases to external sinks. Even when the ECCS of modern nuclear power plants are designed to maintain containment conditions under established limits, there are small amounts of O{sub 2} and H{sub 2} produced mainly by H{sub 2}O radiolysis, and after several days of containment isolation, these gases may accumulate in places where the possibility of a H{sub 2} deflagration increases. Therefore, the use of Pars looks initially as an interesting solution to long-term H{sub 2} control during a Dba. However, the H{sub 2} and O{sub 2} recombination is an exothermic process resulting in additional steam and energy releases to the containment. In order to avoid a direct deposition of the steam and energy (Pars produced) into the containment atmosphere, we propose to use them to operate steam injectors that can bring cold water to replenish ECCS

  11. Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

    Science.gov (United States)

    Leenaers, A.; Joppen, F.; Van den Berghe, S.

    2009-10-01

    In 1975, as a result of a blockage of the coolant inlet flow, two plates of a fuel element of the BR2 reactor of the Belgian Nuclear Research Centre (SCK•CEN) were partially melted. The fuel element consisted of Al-clad plates with 90% 235U enriched UAl x fuel dispersed in an Al matrix. The element had accumulated a burn up of 21% 235U before it was removed from the reactor. Recently, the damaged fuel plates were sent to the hot laboratory for detailed PIE. Microstructural changes and associated temperature markers were used to identify several stages in the progression to fuel melting. It was found that the temperature in the center of the fuel plate had increased above 900-950 °C before the reactor was scrammed. In view of the limited availability of such datasets, the results of this microstructural analysis provide valuable input in the analysis of accident scenarios for research reactors.

  12. Advanced aqueous reprocessing in P and T strategies: process demonstrations on genuine fuels and targets

    Energy Technology Data Exchange (ETDEWEB)

    Satmark, B.; Apostolidis, C.; Courson, O.; Malmbeck, R.; Carlos, R.; Pagliosa, G.; Romer, K.; Glatz, J.P. [European Commission, DG-JRC, Institute for Transuranium Elements, Hot Cell Technology, Karlsruhe (Germany)

    2000-07-01

    In the present work the performance of several processes used for advanced reprocessing of commercial LWR fuels as well as transmutation targets is compared. As a first step uranium and plutonium were recovered by PUREX type reprocessing. The raffinate, containing fission products, lanthanides and the minor actinides (MA) were used as feed for the second step in which minor actinides and lanthanides were separated from the bulk of the fission products. The five different processes tested use CMPO, DIDPA, TRPO, Diamide and CYANEX 923 as extractant. In the third step MA are separated from lanthanides. Here three processes were tested, i.e. using CYANEX 301, the synergistic mixture of di-chloro substituted CYANEX 301 and TOPO, and BTP solvents. Column-, batch- and continuous counter-current extraction techniques were used for the tests. The different processes will be described and discussed in terms of performances and efficiencies for Am and Cm. Efficient separation of MA from different genuine fuel solutions could be demonstrated and thereby also the possibility of closing a future transmutation fuel cycle. The combination, Diamide and BTP was found to be the best among extractants tested to achieve an efficient MA recovery from spent fuel. (authors)

  13. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  14. Advanced aqueous reprocessing in P and T strategies: process demonstrations on genuine fuels and targets

    Energy Technology Data Exchange (ETDEWEB)

    Christiansen, B.; Apostolidis, C.; Carlos, R.; Courson, O.; Glatz, J.P.; Malmbeck, R.; Pagliosa, G.; Roemer, K.; Serrano-Purroy, D. [European Commission, JRC, Inst. for Transuranium Elements, Karlsruhe (Germany)

    2004-07-01

    In the present work the performance of several processes used for advanced reprocessing of commercial LWR fuels as well as transmutation targets is compared. As a first step uranium and plutonium were recovered by PUREX type reprocessing. The raffinate, containing fission products including lanthanides and the minor actinides (MA) was used as feed for the second step in which minor actinides and lanthanides were separated from the bulk of the fission products. The five different processes tested use CMPO, DIDPA, TRPO, diamide and CYANEX 923 as extractants. In the third step MA are separated from lanthanides. Here three processes were tested, i.e. using CYANEX 301, the synergistic mixture of di-chloro substituted CYANEX 301 and TOPO, and BTP solvents. Column-, batch- and continuous counter-current extraction techniques were used for the tests. The different processes will be described and discussed in terms of performances and efficiencies for Am and Cm separation. Efficient separation of MA from different genuine fuel solutions could be demonstrated and thereby also the possibility of closing a future transmutation fuel cycle. The combination of diamide and BTP seems to be the best, among extractants tested, to achieve an efficient MA recovery from spent fuel. (orig.)

  15. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

  16. Melting and Sintering of Ashes

    DEFF Research Database (Denmark)

    Hansen, Lone Aslaug

    1997-01-01

    The thesis contains an experimental study of the fusion and sintering of ashes collected during straw and coal/straw co-firing.A laboratory technique for quantitative determination of ash fusion has been developed based on Simultaneous Thermal Analysis (STA). By means of this method the fraction...... of melt in the investigated ashes has been determined as a function of temperature. Ash fusion results have been correlated to the chemical and mineralogical composition of the ashes, to results from a standard ash fusion test and to results from sintering experiments. Furthermore, the ash fusion results......-firing, the model only had a qualitative agreement with the measured ash deposit formation rates.Sintering measurements were carried out by means of compression strength testing of ash pellets. This method showed to not be applicable for the salt rich fly ash derived from straw combustion. For the fly ashes...

  17. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

  18. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States)

    2010-01-31

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium

  19. Do Melt Inclusions Answer Big Questions?

    Science.gov (United States)

    Hofmann, A. W.; Sobolev, A. V.

    2009-12-01

    In a pioneering paper, Sobolev and Shimizu (1993) demonstrated the existence of ultra-depleted melt inclusions in olivine phenocrysts in MORB. They interpreted these as evidence for the preservation of parental melts formed by progressive near-fractional melting. Subsequently many cases have been described where melt inclusions from single basalt samples display enormous chemical and isotopic heterogeneity. The interpretation of these observations hinges critically on whether such melt inclusions can faithfully preserve primary or parental melt composition. If they do, melt inclusion data can truly answer big questions from small-scale observations. If they do not, they answer rather small questions. Favoring the second possibility, Danyushevsky et al. (2004) have suggested that much of the observed variability of highly incompatible trace elements in melt inclusions “may not represent geologically significant melts, but instead reflect localized, grain-scale reaction processes within the magmatic plumbing system.” We disagree and show that this mechanism cannot, for example, explain isotopic heterogeneity measured in several suites of melt inclusions, nor does it not account for the presence of ultra-depleted melts and "ghost" plagioclase signatures in other inclusions. More recently, Spandler et al. (2007) have suggested on the basis of experimental evidence that diffusion rates for REE in olivine are so rapid that parental melt compositions in melt inclusions are rapidly falsified by diffusional exchange with (evolved) host lava. We show that the very fact that extreme chemical and isotopic heterogeneities are routinely preserved in melt inclusions demonstrates that this conclusion is unwarranted, either because residence times of the olivine phenocrysts are much shorter than assumed by Spandler et al. or because the high experimental diffusion rates are caused by an unknown experimental artifact. Although there is no obvious flaw in design and execution of

  20. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

    The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400) [ ]. Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no

  1. A benchmark initiative on mantle convection with melting and melt segregation

    Science.gov (United States)

    Schmeling, Harro; Dannberg, Juliane; Dohmen, Janik; Kalousova, Klara; Maurice, Maxim; Noack, Lena; Plesa, Ana; Soucek, Ondrej; Spiegelman, Marc; Thieulot, Cedric; Tosi, Nicola; Wallner, Herbert

    2016-04-01

    In recent years a number of mantle convection models have been developed which include partial melting within the asthenosphere, estimation of melt volumes, as well as melt extraction with and without redistribution at the surface or within the lithosphere. All these approaches use various simplifying modelling assumptions whose effects on the dynamics of convection including the feedback on melting have not been explored in sufficient detail. To better assess the significance of such assumptions and to provide test cases for the modelling community we carry out a benchmark comparison. The reference model is taken from the mantle convection benchmark, cases 1a to 1c (Blankenbach et al., 1989), assuming a square box with free slip boundary conditions, the Boussinesq approximation, constant viscosity and Rayleigh numbers of 104 to 10^6. Melting is modelled using a simplified binary solid solution with linearly depth dependent solidus and liquidus temperatures, as well as a solidus temperature depending linearly on depletion. Starting from a plume free initial temperature condition (to avoid melting at the onset time) five cases are investigated: Case 1 includes melting, but without thermal or dynamic feedback on the convection flow. This case provides a total melt generation rate (qm) in a steady state. Case 2 is identical to case 1 except that latent heat is switched on. Case 3 includes batch melting, melt buoyancy (melt Rayleigh number Rm) and depletion buoyancy, but no melt percolation. Output quantities are the Nusselt number (Nu), root mean square velocity (vrms), the maximum and the total melt volume and qm approaching a statistical steady state. Case 4 includes two-phase flow, i.e. melt percolation, assuming a constant shear and bulk viscosity of the matrix and various melt retention numbers (Rt). These cases are carried out using the Compaction Boussinseq Approximation (Schmeling, 2000) or the full compaction formulation. For cases 1 - 3 very good agreement

  2. Bubble Formation in Basalt-like Melts

    DEFF Research Database (Denmark)

    Jensen, Martin; Keding, Ralf; Yue, Yuanzheng

    2011-01-01

    The effect of the melting temperature on bubble size and bubble formation in an iron bearing calcium aluminosilicate melt is studied by means of in-depth images acquired by optical microscopy. The bubble size distribution and the total bubble volume are determined by counting the number of bubble...

  3. Bubble Formation in Basalt-like Melts

    DEFF Research Database (Denmark)

    Jensen, Martin; Keding, Ralf; Yue, Yuanzheng

    2011-01-01

    The effect of the melting temperature on bubble size and bubble formation in an iron bearing calcium aluminosilicate melt is studied by means of in-depth images acquired by optical microscopy. The bubble size distribution and the total bubble volume are determined by counting the number of bubbles...

  4. Disordering and Melting of Aluminum Surfaces

    DEFF Research Database (Denmark)

    Stoltze, Per; Nørskov, Jens Kehlet; Landman, U.

    1988-01-01

    We report on a molecular-dynamics simulation of an Al(110) surface using the effective-medium theory to describe the interatomic interactions. The surface region is found to start melting ≅200 K below the bulk melting temperature with a gradual increase in the thickness of the disordered layer as...

  5. Uniaxial Elongational viscosity of bidisperse polystyrene melts

    DEFF Research Database (Denmark)

    Nielsen, Jens Kromann; Rasmussen, Henrik K.; Hassager, Ole

    2006-01-01

    The startup and steady uniaxial elongational viscosity have been measured for three bidisperse polystyrene (PS) melts, consisting of blends of monodisperse PS with molecular weights of 52 kg/mole or 103 kg/mole and 390 kg/mole. The bidisperse melts have a maximum in the steady elongational...

  6. Uniaxial Elongational viscosity of bidisperse polystyrene melts

    DEFF Research Database (Denmark)

    Nielsen, Jens Kromann; Rasmussen, Henrik K.; Hassager, Ole

    2006-01-01

    The startup and steady uniaxial elongational viscosity have been measured for three bidisperse polystyrene (PS) melts, consisting of blends of monodisperse PS with molecular weights of 52 kg/mole or 103 kg/mole and 390 kg/mole. The bidisperse melts have a maximum in the steady elongational viscos...

  7. Stability of foams in silicate melts

    Science.gov (United States)

    Proussevitch, Alexander A.; Sahagian, Dork L.; Kutolin, Vladislav A.

    1993-12-01

    Bubble coalescence and the spontaneous disruption of high-porosity foams in silicate melts are the result of physical expulsion of interpore melt (syneresis) leading to bubble coalescence, and diffusive gas exchange between bubbles. Melt expulsion can be achieved either along films between pairs of bubbles, or along Plateau borders which represent the contacts between 3 or more bubbles. Theoretical evaluation of these mechanisms is confirmed by experimental results, enabling us to quantify the relevant parameters and determine stable bubble size and critical film thickness in a foam as a function of melt viscosity, surface tension, and time. Foam stability is controlled primarily by melt viscosity and time. Melt transport leading to coalescence of bubbles proceeds along inter-bubble films for smaller bubbles, and along Plateau borders for larger bubbles. Thus the average bubble size accelerates with time. In silicate melts, the diffusive gas expulsion out of a region of foam is effective only for water (and even then, only at small length scales), as the diffusion of CO 2 is negligible. The results of our analyses are applicable to studies of vesicularity of lavas, melt degassing, and eruption mechanisms.

  8. Purification of Niobium by Electron Beam Melting

    Science.gov (United States)

    Sankar, M.; Mirji, K. V.; Prasad, V. V. Satya; Baligidad, R. G.; Gokhale, A. A.

    2016-06-01

    Pure niobium metal, produced by alumino-thermic reduction of niobium oxide, contains various impurities which need to be reduced to acceptable levels to obtain aerospace grade purity. In the present work, an attempt has been made to refine niobium metals by electron beam drip melting technique to achieve purity confirming to the ASTM standard. Input power to the electron gun and melt rate were varied to observe their combined effect on extend of refining and loss of niobium. Electron beam (EB) melting is shown to reduce alkali metals, trace elements and interstitial impurities well below the specified limits. The reduction in the impurities during EB melting is attributed to evaporation and degassing due to the combined effect of high vacuum and high melt surface temperature. The % removal of interstitial impurities is essentially a function of melt rate and input power. As the melt rate decreases or input power increases, the impurity levels in the solidified niobium ingot decrease. The EB refining process is also accompanied by considerable amount of niobium loss, which is attributed to evaporation of pure niobium and niobium sub-oxide. Like other impurities, Nb loss increases with decreasing melt rate or increase in input power.

  9. Shock-induced melting and rapid solidification

    Energy Technology Data Exchange (ETDEWEB)

    Nellis, W.J.; Gourdin, W.H.; Maple, M.B.

    1987-08-01

    Model calculations are presented to estimate that approx.50 GPa is required to completely shock melt metal powders with quenching at rates up to 10/sup 8/ K/s. Experiments are discussed for powders of a Cu-Zr alloy compacted in the usual way at 16 GPa and melted by shocking to 60 GPa. 12 refs.

  10. Validation efforts for the neutronics of a plutonium erbium zirconium oxide inert matrix light water reactor fuel

    Science.gov (United States)

    Paratte, J. M.; Chawla, R.; Früh, R.; Joneja, O. P.; Pelloni, S.; Pralong, C.

    1999-08-01

    Light water reactor (LWR) neutronics codes and cross-section libraries need further qualification when used for the calculation of inert matrix fuel (IMF) cells. Three types of validation efforts have been undertaken for the PuO 2-Er 2O 3-ZrO 2 IMF concept under development at the Paul Scherrer Institute (PSI). Firstly, the PSI calculational scheme, based on the BOXER code and its data library, has been applied to the analysis of a range of LWR experiments with PuO 2-UO 2 fuel, conducted earlier at PSI's PROTEUS facility. The generally good agreement obtained between calculated and measured parameters gives confidence in the ability of the employed calculational scheme to correctly modelize Pu-containing fuel cells. Secondly, reactivity effects of various burnable poisons in a ZrO 2 matrix were measured in the CROCUS reactor of the Swiss Federal Institute of Technology at Lausanne. Modelling these experiments with BOXER resulted in satisfactory prediction of measured reactivity ratios (relative to a soluble-boron standard) for most of the experimental rods employed. This was particularly the case for experiments with erbium, as well as with mixtures of erbium and europium (the latter being used to simulate the effects of overlapping resonances, as would be expected in the case of a Pu-Er IMF). Finally, as there are no experimental results available from power reactors employing IMFs, the validation of burnup calculations (at the cell level) has been based on results obtained in the framework of an international benchmark exercise on the physics of LWRs employing IMFs. Certain discrepancies in calculated parameters have been observed in this context, several of which can be attributed to specific differences in cross-section libraries.

  11. High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G. Robert [Univ. of California, Santa Barbara, CA (United States)

    2017-01-24

    Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences than have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.

  12. Status of the MeLoDIE experiment, an advanced device for the study of the irradiation creep of LWR cladding with full online capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Guimbal, P. [CEA, DEN, DER, Jules Horowitz Reactor, F-13108 St-Paul-lez-Durance (France); Huotilainen, S.; Taehtinen, S. [VTT Technical Research Centre of Finland, Materials performance division, Espoo, FI-02044 VTT (Finland); Thellier, G. [CEA, DEN, DRSN, OSIRIS Reactor, F-91191 Gif-sur-Yvette (France); Villard, J.F. [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, F-13108 St- Paul-lez-Durance (France)

    2015-07-01

    As a prototype of future instrumented material experiments in the Jules Horowitz Reactor (JHR), the MELODIE project was launched in 2009 by the CEA in collaboration with VTT. Being designed as a biaxial creep experiment with online capability, MELODIE is able to apply an online-controlled biaxial loading on a LWR clad sample up to 120 MPa and to perform an online measurement of its biaxial deformation. An important experimental challenge was to perform reliably accurate measurements under the high nuclear heat load of in-core locations while keeping within their tight space. For that purpose, specific sensors were co-designed with and built by IFE Halden. Manufacturing of the MELODIE components was completed one year ago. The complexity of its in-pile section and of the pressurization system requested a step-by-step tuning of the setup. The toughest part of this process dealt with the Diameter gauge which required a partial redesign to take into account unexpected and unwanted electromagnetic interactions with the hosting device. Final cold performance tests of the on-board instrumentation will be presented. The MELODIE device is now ready and irradiation should start in OSIRIS reactor this spring. (authors)

  13. Long-term embrittlement of cast duplex stainless steels in LWR systems. Semiannual report, October 1991--March 1992: Volume 7, No.1

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K. [Argonne National Lab., IL (United States)

    1993-05-01

    This progress report summarizes work performed by Argonne National Laboratory on long-term thermal embrittlement of cast duplex stainless steels in LWR systems during the six months from October 1991 to March 1992. Charpy-impact, tensile, and fracture toughness J-R curve data are presented for several heats of cast stainless steel that were aged 10,000-58,000 h at 290, 320, and 350{degree}C. The results indicate that thermal aging decreases the fracture toughness of cast stainless steels. In general, CF-3 steels are the least sensitive to thermal aging and CF-8M steels are the most sensitive. The values of fracture toughness J{sub IC} and tearing modulus for CF-8M steels can be as low as {approx}90 kJ/m{sup 2} and {approx}60, respectively. The fracture toughness data are consistent with the Charpy-impact results, i.e. unaged and aged steels that show low impact energy also exhibit lower fracture toughness. All steels reach a minimum saturation fracture toughness after thermal aging; the time to reach saturation depends on the aging temperature. The results also indicate that low-strength cast stainless steels are generally insensitive to thermal aging.

  14. Aviation fuels outlook

    Science.gov (United States)

    Momenthy, A. M.

    1980-01-01

    Options for satisfying the future demand for commercial jet fuels are analyzed. It is concluded that the most effective means to this end are to attract more refiners to the jet fuel market and encourage development of processes to convert oil shale and coal to transportation fuels. Furthermore, changing the U.S. refineries fuel specification would not significantly alter jet fuel availability.

  15. Viscosity of carbonate-rich melts under different oxygen fugacity conditions

    Science.gov (United States)

    Di Genova, Danilo; Hess, Kai-Uwe; Cimarelli, Corrado; Dingwell, Donald B.

    2015-04-01

    Viscosity is a fundamental property of many materials and its changes affects the fluid dynamics of natural system as well as industrial processes. The mobility of carbonatitic melts, which are carbonate-rich and very fluid melts, has attracted renewed interest in both earth science and industry. In fact, these melts are considered the main transport agent of carbon from the mantle to the crust and may be intimately linked to the generation of kimberlites. At the same time lithium, potassium and sodium carbonate are used as electrolytes in molten carbonate fuel cells which operate at high temperatures (~650° C) for the production of electricity without CO2 emissions. Accurate measurement of the transport property (i.e. viscosity) of carbonatitic melts is a priority in order to understand the carbonatite mobility and reaction rates. Additionally, obtaining accurate viscosity measurements of such low viscosity melts is however an experimental challenge due to volatility, very low torques and chemical melt instability in the viscometer. To overcome these limitations we have customized a Modular Compact Rheometer (MCR 502 from Anton Paar) ad hoc equipped with 2 narrow gap concentric-cylinder geometries of steel and Pt-Au. The rheometer is characterized by an air-bearing-supported synchronous motor with torque ranging between 0.01 μNm and 230 mNm (resolution of 0.1 nNm), achieving very low viscosity measurements in the order of mPa s, temperatures up to 1000° C and shear rates ranging between 1 and 100 sec-1. These experimental conditions well match the temperature-viscosity-shear rate window relevant for carbonate melts. Here we present the calibration of the rheometer and the results of a rheological characterization study on a series of very low viscous synthetic and natural carbonatitic melts at different oxygen fugacity (air and CO2 saturated atmosphere). Viscosity measurements on carbonate melts have been performed in the temperature range between ~650 and 1000

  16. Size-dependent melting of Bi nanoparticles

    Science.gov (United States)

    Olson, E. A.; Efremov, M. Yu.; Zhang, M.; Zhang, Z.; Allen, L. H.

    2005-02-01

    Nanocalorimetry was used to investigate the melting of Bi nanoparticles. The particles were formed by evaporating Bi onto a silicon nitride substrate, which was then heated. The particles self-assemble into truncated spherical particles. Below 5-nm average film thickness, mean particle sizes increased linearly with deposition thickness but increased rapidly for 10-nm-thick films. As expected, small particles were found to exhibit size-dependent melting temperatures less than the bulk melting temperature (e.g., ΔT =67K for a 3-nm radius particle). The measured melting temperatures for particles below ˜7nm in radius, however, were ˜50K above the value predicted by the homogeneous melting model. We discuss this discrepancy in terms of a possible size-dependent crystal structure change and the superheating of the solid phase.

  17. Nanotexturing of surfaces to reduce melting point.

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, Ernest J.; Zubia, David (University of Texas at El Paso El Paso, TX); Mireles, Jose (Universidad Aut%C3%94onoma de Ciudad Ju%C3%94arez Ciudad Ju%C3%94arez, Mexico); Marquez, Noel (University of Texas at El Paso El Paso, TX); Quinones, Stella (University of Texas at El Paso El Paso, TX)

    2011-11-01

    This investigation examined the use of nano-patterned structures on Silicon-on-Insulator (SOI) material to reduce the bulk material melting point (1414 C). It has been found that sharp-tipped and other similar structures have a propensity to move to the lower energy states of spherical structures and as a result exhibit lower melting points than the bulk material. Such a reduction of the melting point would offer a number of interesting opportunities for bonding in microsystems packaging applications. Nano patterning process capabilities were developed to create the required structures for the investigation. One of the technical challenges of the project was understanding and creating the specialized conditions required to observe the melting and reshaping phenomena. Through systematic experimentation and review of the literature these conditions were determined and used to conduct phase change experiments. Melting temperatures as low as 1030 C were observed.

  18. Volatile diffusion in silicate melts and its effects on melt inclusions

    Directory of Open Access Journals (Sweden)

    P. Scarlato

    2005-06-01

    Full Text Available A compendium of diffusion measurements and their Arrhenius equations for water, carbon dioxide, sulfur, fluorine, and chlorine in silicate melts similar in composition to natural igneous rocks is presented. Water diffusion in silicic melts is well studied and understood, however little data exists for melts of intermediate to basic compositions. The data demonstrate that both the water concentration and the anhydrous melt composition affect the diffusion coefficient of water. Carbon dioxide diffusion appears only weakly dependent, at most, on the volatilefree melt composition and no effect of carbon dioxide concentration has been observed, although few experiments have been performed. Based upon one study, the addition of water to rhyolitic melts increases carbon dioxide diffusion by orders of magnitude to values similar to that of 6 wt% water. Sulfur diffusion in intermediate to silicic melts depends upon the anhydrous melt composition and the water concentration. In water-bearing silicic melts sulfur diffuses 2 to 3 orders of magnitude slower than water. Chlorine diffusion is affected by both water concentration and anhydrous melt composition; its values are typically between those of water and sulfur. Information on fluorine diffusion is rare, but the volatile-free melt composition exerts a strong control on its diffusion. At the present time the diffusion of water, carbon dioxide, sulfur and chlorine can be estimated in silicic melts at magmatic temperatures. The diffusion of water and carbon dioxide in basic to intermediate melts is only known at a limited set of temperatures and compositions. The diffusion data for rhyolitic melts at 800°C together with a standard model for the enrichment of incompatible elements in front of growing crystals demonstrate that rapid crystal growth, greater than 10-10 ms-1, can significantly increase the volatile concentrations at the crystal-melt interface and that any of that melt trapped

  19. Melt Rate Improvement for DWPF MB3: Melt Rate Furnace Testing

    Energy Technology Data Exchange (ETDEWEB)

    Stone, M.E.

    2001-07-24

    The Defense Waste Processing Facility (DWPF) would like to increase its canister production rate. The goal of this study is to improve the melt rate in DWPF specifically for Macrobatch 3. However, the knowledge gained may result in improved melting efficiencies translating to future DWPF macrobatches and in higher throughput for other Department of Energy's (DOE) melters. Increased melting efficiencies decrease overall operational costs by reducing the immobilization campaign time for a particular waste stream. For melt rate limited systems, a small increase in melting efficiency translates into significant hard dollar savings by reducing life cycle operational costs.

  20. Nuclear spent fuel management scenarios. Status and assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Dufek, J.; Arzhanov, V.; Gudowski, W. [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Nuclear and Reactor Physics

    2006-06-15

    The strategy for management of spent nuclear fuel from the Swedish nuclear power programme is interim storage for cooling and decay for about 30 years followed by direct disposal of the fuel in a geologic repository. In various contexts it is of interest to compare this strategy with other strategies that might be available in the future as a result of ongoing research and development. In particular partitioning and transmutation is one such strategy that is subject to considerable R and D-efforts within the European Union and in other countries with large nuclear programmes. To facilitate such comparisons for the Swedish situation, with a planned phase out of the nuclear power programme, SKB has asked the team at Royal Inst. of Technology to describe and explore some scenarios that might be applied to the Swedish programme. The results of this study are presented in this report. The following scenarios were studied by the help of a specially developed computer programme: Phase out by 2025 with direct disposal. Burning plutonium and minor actinides as MOX in BWR. Burning plutonium and minor actinides as MOX in PWR. Burning plutonium and minor actinides in ADS. Combined LWR-MOX plus ADS. For the different scenarios nuclide inventories, waste amounts, costs, additional electricity production etc have been assessed. As a general conclusion it was found that BWR is more efficient for burning plutonium in MOX fuel than PWR. The difference is approximately 10%. Furthermore the BWR produces about 10% less americium inventory. An ADS reactor park can theoretically in an ideal case burn (transmute) 99% of the transuranium isotopes. The duration of such a scenario heavily depends on the interim time needed for cooling the spent fuel before reprocessing. Assuming 10 years for cooling of nuclear fuel from ADS, the duration will be at least 200 years under optimistic technical assumptions. The development and use of advanced pyro-processing with an interim cooling time of only

  1. Partitioning coefficients between olivine and silicate melts

    Science.gov (United States)

    Bédard, J. H.

    2005-08-01

    Variation of Nernst partition coefficients ( D) between olivine and silicate melts cannot be neglected when modeling partial melting and fractional crystallization. Published natural and experimental olivine/liquidD data were examined for covariation with pressure, temperature, olivine forsterite content, and melt SiO 2, H 2O, MgO and MgO/MgO + FeO total. Values of olivine/liquidD generally increase with decreasing temperature and melt MgO content, and with increasing melt SiO 2 content, but generally show poor correlations with other variables. Multi-element olivine/liquidD profiles calculated from regressions of D REE-Sc-Y vs. melt MgO content are compared to results of the Lattice Strain Model to link melt MgO and: D0 (the strain compensated partition coefficient), EM3+ (Young's Modulus), and r0 (the size of the M site). Ln D0 varies linearly with Ln MgO in the melt; EM3+ varies linearly with melt MgO, with a dog-leg at ca. 1.5% MgO; and r0 remains constant at 0.807 Å. These equations are then used to calculate olivine/liquidD for these elements using the Lattice Strain Model. These empirical parameterizations of olivine/liquidD variations yield results comparable to experimental or natural partitioning data, and can easily be integrated into existing trace element modeling algorithms. The olivine/liquidD data suggest that basaltic melts in equilibrium with pure olivine may acquire small negative Ta-Hf-Zr-Ti anomalies, but that negative Nb anomalies are unlikely to develop. Misfits between results of the Lattice Strain Model and most light rare earth and large ion lithophile partitioning data suggest that kinetic effects may limit the lower value of D for extremely incompatible elements in natural situations characterized by high cooling/crystallization rates.

  2. Incorporation of Integral Fuel Burnable Absorbers Boron and Gadolinium into Zirconium-Alloy Fuel Clad Material

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, K.; Renk, T.J.; Lahoda, E.J.; Corradini, M.L

    2004-12-14

    Long-lived fuels require the use of higher enrichments of 235U or other fissile materials. Such high levels of fissile material lead to excessive fuel activity at the beginning of life. To counteract this excessive activity, integral fuel burnable absorbers (IFBA) are added to some rods in the fuel assembly. The two commonly used IFBA elements are gadolinium, which is added as gadolinium-oxide to the UO2 powder, and boron, which is applied as a zirconium-diboride coating on the UO2 pellets using plasma spraying or chemical vapor deposition techniques. The incorporation of IFBA into the fuel has to be performed in a nuclear-regulated facility that is physically separated from the main plant. These operations tend to be very costly because of their small volume and can add from 20 to 30% to the manufacturing cost of the fuel. Other manufacturing issues that impact cost and performance are maintaining the correct levels of dosing, the reduction in fuel melting point due to gadolinium-oxide additions, and parasitic neutron absorpt