Science.gov

Sample records for core burnup calculations

  1. Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations

    SciTech Connect

    Wagner, J.C.; DeHart, M.D.

    2000-03-01

    This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ``end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified.

  2. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    SciTech Connect

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  3. High-burnup core design using minor actinide-containing metal fuel

    SciTech Connect

    Ohta, Hirokazu; Ogata, Takanari; Obara, T.

    2013-07-01

    A neutronic design study of metal fuel fast reactor (FR) cores is conducted on the basis of an innovative fuel design concept to achieve an extremely high burnup and realize an efficient fuel cycle system. Since it is expected that the burnup reactivity swing will become extremely large in an unprecedented high burnup core, minor actinides (MAs) from light water reactors (LWRs) are added to fresh fuel to improve the core internal conversion. Core neutronic analysis revealed that high burnups of about 200 MWd/kg for a small-scale core and about 300 MWd/kg for a large-scale core can be attained while suppressing the burnup reactivity swing to almost the same level as that of conventional cores with normal burnup. An actinide burnup analysis has shown that the MA consumption ratio is improved to about 60% and that the accumulated MAs originating from LWRs can be efficiently consumed by the high-burnup metal fuel FR. (authors)

  4. PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

    DOEpatents

    Puechl, K.H.

    1963-09-24

    A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

  5. MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION

    SciTech Connect

    Nichols, T.; Sternat, M.; Charlton, W.

    2011-05-08

    MONTEBURNS is a Monte-Carlo depletion routine utilizing MCNP and ORIGEN 2.2. Uncertainties exist in the MCNP transport calculation, but this information is not passed to the depletion calculation in ORIGEN or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of a multiple repeated depletion runs is performed. The reactor model chosen is the Oak Ridge Research Reactor (ORR) in a single assembly, infinite lattice configuration. This model was burned for a 25.5 day cycle broken down into three steps. The output isotopics as well as effective multiplication factor (k-effective) were tabulated and histograms were created at each burnup step using the Scott Method to determine the bin width. It was expected that the gram quantities and k-effective histograms would produce normally distributed results since they were produced from a Monte-Carlo routine, but some of results do not. The standard deviation at each burnup step was consistent between fission product isotopes as expected, while the uranium isotopes created some unique results. The variation in the quantity of uranium was small enough that, from the reaction rate MCNP tally, round off error occurred producing a set of repeated results with slight variation. Statistical analyses were performed using the {chi}{sup 2} test against a normal distribution for several isotopes and the k-effective results. While the isotopes failed to reject the null hypothesis of being normally distributed, the {chi}{sup 2} statistic grew through the steps in the k-effective test. The null hypothesis was rejected in the later steps. These results suggest, for a high accuracy solution, MCNP cell material quantities less than 100 grams and greater kcode parameters are needed to minimize uncertainty propagation and minimize round off effects.

  6. Burnup concept for a long-life fast reactor core using MCNPX.

    SciTech Connect

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  7. Comparison of PWR - Burnup calculations with SCALE 5.0/TRITON other burnup codes and experimental results

    SciTech Connect

    Oberle, P.; Broeders, C. H. M.; Dagan, R.

    2006-07-01

    The increasing tendency towards fuel lifetime extension in thermal nuclear reactors motivated validation work for available evaluation tools for nuclear fuel burnup calculations. In this study two deterministic codes with different transport solvers and one Monte Carlo method are investigated. The code system KAPROS/KARBUS uses the classical deterministic First Collision Probability method utilizing a cylinderized Wigner-Seitz cell. In the SCALES.0/TRITON/NEWT code the Extended Step Characteristic method is applied. In a first step the two deterministic codes are compared with experimental results from the KWO-Isotope Correlation Experiment up to 30 MWD/kg HM burnup, published in 1981. Two pin cell calculations are analyzed by comparison of calculated and experimental results for important heavy isotope vectors. The results are very satisfactory. Subsequently, further validation at higher burnup (< 80 MWD/kg HM) is provided by comparison of the two deterministic codes and the Monte Carlo based burnup code MONTEBURNS for PWR UO{sub 2} fuel assembly calculations. Possible reasons for differences in the results are analyzed and discussed. Especially the influence of cross section data and processing is presented. (authors)

  8. Accident source terms for boiling water reactors with high burnup cores.

    SciTech Connect

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  9. Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

    SciTech Connect

    Mueller, Don; Bowen, Douglas G; Marshall, William BJ J

    2015-01-01

    The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members accept the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δkeff (ISG-8, Rev. 3, Recommendation 4).

  10. BOXER: Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations

    2010-02-01

    Neutron transport, calculation of multiplication factor and neutron fluxes in 2-D configurations: cell calculations, 2-D diffusion and transport, and burnup. Preparation of a cross section library for the code BOXER from a basic library in ENDF/B format (ETOBOX).

  11. Fuel rod and core materials investigations related to LWR extended burnup operation

    NASA Astrophysics Data System (ADS)

    Kolstad, Erik; Vitanza, Carlo

    1992-06-01

    The paper deals with tests and recent measurements related to extended burnup fuel performance and describes test facilities and results in the areas of waterside cladding corrosion and irradiation-assisted stress corrosion cracking (IASCC). Fuel temperature data suggest a gradual degradation of UO 2 thermal conductivity with exposure in the range 6-8% per 10 MWd/kgUO 2 at temperatures below 700°C. The effect on the fuel microstructure of interlinkage and resintering phenomena is shown by measuring the surface-to-volume ( S/ V) ratio of the fuel. Changes in S/V with burnup are correlated to power rating and fuel operating temperature. No evidence was found of enhanced fission gas release during load-follow operation in the burnup range 25-45 MWd/kgUO 2. The effect of high lithium concentration (high pH) on the corrosion behaviour of pre-irradiated high burnup Zircaloy-4 fuel rods subjected either to nucleate boiling or to one-phase cooling conditions was studied. The oxide thickness growth rates measured at an average burnup up to 40 MWd/kgUO 2 are consistent with literature data and show no evidence of corrosion enhancement due to the high lithium content and little effect of cooling regime. A test facility for exploring the effects of environmental variables on IASCC behaviour of in-core structural materials is described.

  12. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    NASA Astrophysics Data System (ADS)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  13. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    SciTech Connect

    Afifah, Maryam Su’ud, Zaki; Miura, Ryosuke; Takaki, Naoyuki; Sekimoto, H.

    2015-09-30

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  14. Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

    SciTech Connect

    Kucukboyaci, Vefa; Marshall, William BJ J

    2012-01-01

    In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse's state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.

  15. Burnup calculation by the method of first-flight collision probabilities using average chords prior to the first collision

    NASA Astrophysics Data System (ADS)

    Karpushkin, T. Yu.

    2012-12-01

    A technique to calculate the burnup of materials of cells and fuel assemblies using the matrices of first-flight neutron collision probabilities rebuilt at a given burnup step is presented. A method to rebuild and correct first collision probability matrices using average chords prior to the first neutron collision, which are calculated with the help of geometric modules of constructed stochastic neutron trajectories, is described. Results of calculation of the infinite multiplication factor for elementary cells with a modified material composition compared to the reference one as well as calculation of material burnup in the cells and fuel assemblies of a VVER-1000 are presented.

  16. OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

    SciTech Connect

    DeHart, M.D.

    1993-01-01

    Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are {sup 149}Sm, {sup 151}Sm, and {sup 155}Gd.

  17. OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

    SciTech Connect

    DeHart, M.D.; Parks, C.V.; Brady, M.C.

    1996-06-01

    In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155.

  18. Large Core Code Evaluation Working Group Benchmark Problem Four: neutronics and burnup analysis of a large heterogeneous fast reactor. Part 1. Analysis of benchmark results. [LMFBR

    SciTech Connect

    Cowan, C.L.; Protsik, R.; Lewellen, J.W.

    1984-01-01

    The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the techniques for modeling the three-dimensional benchmark geometry, and sensitivity studies were carried out to determine the performance parameter sensitivities to changes in the neutronics and burnup specifications. The results of the Benchmark Four calculations indicated that a linked RZ-XY (Hex) two-dimensional representation of the benchmark model geometry can be used to predict mass balance data, power distributions, regionwise fuel exposure data and burnup reactivities with good accuracy when compared with the results of direct three-dimensional computations. Most of the small differences in the results of the benchmark analyses by the different participants were attributed to ambiguities in carrying out the regionwise flux renormalization calculations throughout the burnup step.

  19. Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

    SciTech Connect

    Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

    2005-05-24

    The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

  20. Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

    NASA Astrophysics Data System (ADS)

    Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

    2005-05-01

    The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

  1. Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations

    DOE PAGES

    Fensin, M. L.; Galloway, J. D.; James, M. R.

    2015-04-11

    The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and newmore » predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.« less

  2. Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations

    SciTech Connect

    Fensin, M. L.; Galloway, J. D.; James, M. R.

    2015-04-11

    The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and new predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.

  3. Propagation of Uncertainty in System Parameters of a LWR Model by Sampling MCNPX Calculations - Burnup Analysis

    NASA Astrophysics Data System (ADS)

    Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.

    2014-06-01

    For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.

  4. Burnup calculations and chemical analysis of irradiated fuel samples studied in LWR-PROTEUS phase II

    SciTech Connect

    Grimm, P.; Guenther-Leopold, I.; Berger, H. D.

    2006-07-01

    The isotopic compositions of 5 UO{sub 2} samples irradiated in a Swiss PWR power plant, which were investigated in the LWR-PROTEUS Phase II programme, were calculated using the CASMO-4 and BOXER assembly codes. The burnups of the samples range from 50 to 90 MWd/kg. The results for a large number of actinide and fission product nuclides were compared to those of chemical analyses performed using a combination of chromatographic separation and mass spectrometry. A good agreement of calculated and measured concentrations is found for many of the nuclides investigated with both codes. The concentrations of the Pu isotopes are mostly predicted within {+-}10%, the two codes giving quite different results, except for {sup 242}Pu. Relatively significant deviations are found for some isotopes of Cs and Sm, and large discrepancies are observed for Eu and Gd. The overall quality of the predictions by the two codes is comparable, and the deviations from the experimental data do not generally increase with burnup. (authors)

  5. Calculated Neutron and Gamma-ray Spectra across the Prismatic Very High Temperature Reactor Core

    SciTech Connect

    James W. Sterbentz

    2008-05-01

    Neutron and gamma-ray flux spectra are calculated using the MCNP5 computer code and a one-sixth core model of a prismatic Very High Temperature Reactor based on the General Atomics Gas Turbine-Modular Helium Reactor. Spectra are calculated in the five inner reflector graphite block rings, three annular active core fuel rings, three outer graphite reflector block rings, and the core barrel. The neutron spectra are block and fuel pin averages and are calculated as a function of temperature and burnup. Also provided are the total, fast, and thermal radial profile fluxes and core barrel dpa rates.

  6. Substantiation of parameters of the geometric model of the research reactor core for the calculation using the Monte Carlo method

    SciTech Connect

    Radaev, A. I. Schurovskaya, M. V.

    2015-12-15

    The choice of the spatial nodalization for the calculation of the power density and burnup distribution in a research reactor core with fuel assemblies of the IRT-3M and VVR-KN type using the program based on the Monte Carlo code is described. The influence of the spatial nodalization on the results of calculating basic neutronic characteristics and calculation time is investigated.

  7. Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor.

    2001-08-29

    Version 00 TRIGLAV is a computer program for reactor calculations of mixed cores in a TRIGA Mark II research reactor. It can be applied for fuel element burn-up calculations, for power and flux distributions calculations and for reactivity predictions. The TRIGLAV program requires the WIMS-D4 program with the original WIMS cross-section library extended for TRIGA reactor specific nuclides. This package includes the code TRIGAC, which is a new version of TRIGAP.

  8. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    NASA Astrophysics Data System (ADS)

    Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

    2013-10-01

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  9. Propagation of Nuclear Data Uncertainties for ELECTRA Burn-up Calculations

    NASA Astrophysics Data System (ADS)

    Sjöstrand, H.; Alhassan, E.; Duan, J.; Gustavsson, C.; Koning, A. J.; Pomp, S.; Rochman, D.; Österlund, M.

    2014-04-01

    The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in 239Pu transport data to uncertainties in the fuel inventory of ELECTRA during the reactor lifetime using the Total Monte Carlo approach (TMC). Within the TENDL project, nuclear models input parameters were randomized within their uncertainties and 740 239Pu nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long-term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty of some minor actinides were observed to be rather large and therefore their impact on multiple recycling should be investigated further. It was also found that, criticality benchmarks can be used to reduce inventory uncertainties due to nuclear data. Further studies are needed to include fission yield uncertainties, more isotopes, and a larger set of benchmarks.

  10. Development of Monteburns: A Code That Links MCNP and ORIGEN2 in an Automated Fashion for Burnup Calculations

    SciTech Connect

    Holly R. Trellue

    1998-12-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

  11. Burnup calculations for KIPT accelerator driven subcritical facility using Monte Carlo computer codes-MCB and MCNPX.

    SciTech Connect

    Gohar, Y.; Zhong, Z.; Talamo, A.; Nuclear Engineering Division

    2009-06-09

    Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is {approx}375 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the

  12. Conservative axial burnup distributions for actinide-only burnup credit

    SciTech Connect

    Kang, C.; Lancaster, D.

    1997-11-01

    Unlike the fresh fuel approach, which assumes the initial isotopic compositions for criticality analyses, any burnup credit methodology must address the proper treatment of axial burnup distributions. A straightforward way of treating a given axial burnup distribution is to segment the fuel assembly into multiple meshes and to model each burnup mesh with the corresponding isotopic compositions. Although this approach represents a significant increase in modeling efforts compared to the uniform average burnup approach, it can adequately determine the reactivity effect of the axial burnup distribution. A major consideration is what axial burnup distributions are appropriate for use in light of many possible distributions depending on core operating conditions and histories. This paper summarizes criticality analyses performed to determine conservative axial burnup distributions. The conservative axial burnup distributions presented in this paper are included in the Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages, Revision 1 submitted in May 1997 by the US Department of Energy (DOE) to the US Nuclear Regulatory Commission (NRC). When approved by NRC, the conservative axial burnup distributions may be used to model PWR spent nuclear fuel for the purpose of gaining actinide only burnup credit.

  13. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    NASA Astrophysics Data System (ADS)

    Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay

    2016-02-01

    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  14. IN-CORE FUEL MANAGEMENT: PWR Core Calculations Using MCRAC

    NASA Astrophysics Data System (ADS)

    PetroviĆ, B. G.

    1991-01-01

    The following sections are included: * INTRODUCTION * IN-CORE FUEL MANAGEMENT CALCULATIONS * In-Core Fuel Management * Methodological Problems of In-Core Fuel Management * In-Core Fuel Management Analytical Tools * PENN STATE FUEL MANAGEMENT PACKAGE * Penn State Fuel Management Package (PFMP) * Assembly Data Description (ADD) * Linking PSU-LEOPARD and MCRAC: An Example * MULTICYCLE REACTOR ANALYSIS CODE (MCRAC) * Main Features and Options of MCRAC code * Core geometry * Diffusion equations * 1.5-group model * Multicycle neutronic analysis * Multicycle cost analysis * Criticality search * Power-dependent xenon feedback calculations * Control rod and burnable absorber simulation * Search for LP with flat BOC power distribution * Artificial ADD option * Variable dimensioning technique * RBI version of MCRAC code * Programming changes in PC version * Fuel interchange option * MCRAC Input/Output * General input description * Sample input * Sample output * EXPERIENCE WITH MCRAC CODE * CONCLUSIONS * REFERENCES

  15. Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations

    SciTech Connect

    Fensin, Michael Lorne; Umbel, Marissa

    2015-09-18

    Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fission yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.

  16. Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations

    DOE PAGES

    Fensin, Michael Lorne; Umbel, Marissa

    2015-09-18

    Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fissionmore » yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.« less

  17. Power excursion analysis for BWR`s at high burnup

    SciTech Connect

    Diamond, D.J.; Neymoith, L.; Kohut, P.

    1996-03-01

    A study has been undertaken to determine the fuel enthalpy during a rod drop accident and during two thermal-hydraulic transients. The objective was to understand the consequences to high burnup fuel and the sources of uncertainty in the calculations. The analysis was done with RAMONA-4B, a computer code that models the neutron kinetics throughout the core along with the thermal-hydraulics in the core, vessel, and steamline. The results showed that the maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important parameters in each of these categories are discussed in the paper.

  18. SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

    SciTech Connect

    BSC

    2004-12-01

    Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier

  19. Appropriate burnup measurements for transportation burnup credit

    SciTech Connect

    Lancaster, D.; Fuentes, E.

    1997-04-01

    This paper addresses two of the measurement specifications used in analyzing spent fuel packages to gain burnup credit. The philosophy and calculation of rejection criteria and measurement accuracy are discussed. Any assembly for which the declared measured value and reactor record value deviate by more than 10% will be rejected. Measurement accuracy requirements are established for dependent and independent systems. The requirements have been tested and are achievable, ensuring safe operation without extra cost. 6 refs.

  20. Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

    SciTech Connect

    Schneider, E.; Deinert, M.; Bingham Cady, K.

    2006-07-01

    Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

  1. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

  2. Dose Rate Calculations for Rotary Mode Core Sampling Exhauster

    SciTech Connect

    FOUST, D.J.

    2000-10-26

    This document provides the calculated estimated dose rates for three external locations on the Rotary Mode Core Sampling (RMCS) exhauster HEPA filter housing, per the request of Characterization Field Engineering.

  3. Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Butler, C.; Albright, D.

    2007-01-01

    Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

  4. FRAPCON-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup

    SciTech Connect

    Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.

    1997-12-01

    FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. 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 phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes` integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6).

  5. Modifying scoping codes to accurately calculate TMI-cores with lifetimes greater than 500 effective full-power days

    SciTech Connect

    Bai, D.; Levine, S.L. ); Luoma, J.; Mahgerefteh, M. )

    1992-01-01

    The Three Mile Island unit 1 core reloads have been designed using fast but accurate scoping codes, PSUI-LEOPARD and ADMARC. PSUI-LEOPARD has been normalized to EPRI-CPM2 results and used to calculate the two-group constants, whereas ADMARC is a modern two-dimensional, two-group diffusion theory nodal code. Problems in accuracy were encountered for cycles 8 and higher as the core lifetime was increased beyond 500 effective full-power days. This is because the heavier loaded cores in both {sup 235}U and {sup 10}B have harder neutron spectra, which produces a change in the transport effect in the baffle reflector region, and the burnable poison (BP) simulations were not accurate enough for the cores containing the increased amount of {sup 10}B required in the BP rods. In the authors study, a technique has been developed to take into account the change in the transport effect in the baffle region by modifying the fast neutron diffusion coefficient as a function of cycle length and core exposure or burnup. A more accurate BP simulation method is also developed, using integral transport theory and CPM2 data, to calculate the BP contribution to the equivalent fuel assembly (supercell) two-group constants. The net result is that the accuracy of the scoping codes is as good as that produced by CASMO/SIMULATE or CPM2/SIMULATE when comparing with measured data.

  6. Calculation on flux-MMF relationship of orthogonal-core

    SciTech Connect

    Tajima, K.; Kaga, A.; Anazawa, Y. ); Ichinokura, O. )

    1993-03-01

    Orthogonal-cores have various potential applications, for instance in parametric transformers and dc-ac converters. The operating characteristics of the devices can be calculated on the basis of the measured relationship of flux to MMF of the orthogonal-core. To achieve optimal design of the applied device, the relationship of flux to MMF must be determined; however, this involves solving a three dimensional nonlinear problem. In this paper, the authors calculate the flux-MMF relationship based on a magnetic circuit model for the orthogonal-core. The computed results agree well with experiment. The method of this study is shown to be valid for calculation of characteristics and useful for optimal design of application devices.

  7. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    SciTech Connect

    Mueller, Don E.; Marshall, William J.; Wagner, John C.; Bowen, Douglas G.

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  8. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    SciTech Connect

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria

  9. Value of burnup credit beyond actinides

    SciTech Connect

    Lancaster, D.; Fuentes, E.; Kang, Chi

    1997-12-01

    DOE has submitted a topical report to the NRC justifying burnup credit based only on actinide isotopes (U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel to be transported in significantly higher capacity casks. A cost savings estimate for shipping fuel in 32 assembly (burnup credit) casks as opposed to 24 assembly (non-burnup credit) casks was previously presented. Since that time, more detailed calculations have been performed using the methodology presented in the Actinide-Only Burnup Credit Topical Report. Loading curves for derated casks have been generated using actinide-only burnup credit and are presented in this paper. The estimates of cost savings due to burnup credit for shipping fuel utilizing 32, 30, 28, and 24 assembly casks where only the 24 assembly cask does not burnup credit have been created and are discussed. 4 refs., 2 figs.

  10. Spatial Kinetics Calculations of MOX Fueled Core: Variant 22

    SciTech Connect

    Pavlovichev, A.M.

    2001-01-11

    This work is part of a Joint US/Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactors and presents the results of spatial kinetics calculational benchmarks. The examinations were carried out with the following purposes: to verify one of spatial neutronic kinetics model elaborated in KI, to understand sensibility of the model to neutronics difference of UOX and MOX cores, to compare in future point and spatial kinetics models (on the base of a set of selected accidents) in view of eventual creation of RELAP option with 3D kinetics. The document contains input data and results of model operation of three emergency dynamic processes in the VVER-1000 core: central control rod ejection by pressure drop caused by destroying of the moving mechanism cover; overcooling of the reactor core caused by steam line rupture and non-closure of steam generator stop valve; and the boron dilution of coolant in part of the VVER-1000 core caused by penetration of the distillate slug into the core at start up of non-working loop.

  11. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    NASA Astrophysics Data System (ADS)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

  12. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    SciTech Connect

    Marshall, William BJ J; Ade, Brian J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Mertyurek, Ugur; Radulescu, Georgeta

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  13. Recent Developments in No-Core Shell-Model Calculations

    SciTech Connect

    Navratil, P; Quaglioni, S; Stetcu, I; Barrett, B R

    2009-03-20

    We present an overview of recent results and developments of the no-core shell model (NCSM), an ab initio approach to the nuclear many-body problem for light nuclei. In this aproach, we start from realistic two-nucleon or two- plus three-nucleon interactions. Many-body calculations are performed using a finite harmonic-oscillator (HO) basis. To facilitate convergence for realistic inter-nucleon interactions that generate strong short-range correlations, we derive effective interactions by unitary transformations that are tailored to the HO basis truncation. For soft realistic interactions this might not be necessary. If that is the case, the NCSM calculations are variational. In either case, the ab initio NCSM preserves translational invariance of the nuclear many-body problem. In this review, we, in particular, highlight results obtained with the chiral two- plus three-nucleon interactions. We discuss efforts to extend the applicability of the NCSM to heavier nuclei and larger model spaces using importance-truncation schemes and/or use of effective interactions with a core. We outline an extension of the ab initio NCSM to the description of nuclear reactions by the resonating group method technique. A future direction of the approach, the ab initio NCSM with continuum, which will provide a complete description of nuclei as open systems with coupling of bound and continuum states is given in the concluding part of the review.

  14. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  15. Impact investigation of reactor fuel operating parameters on reactivity for use in burnup credit applications

    NASA Astrophysics Data System (ADS)

    Sloma, Tanya Noel

    When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light

  16. Isotopic validation for PWR actinide-only burnup credit using Yankee Rowe data

    SciTech Connect

    1997-11-01

    Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ``fresh fuel`` assumption is very conservative since the potential reactivity of the nuclear fuel is substantially reduced after being irradiated in the reactor core. The concept of taking credit for this reduction in nuclear fuel reactivity due to burnup of the fuel, instead of using the fresh fuel assumption in the criticality safety analysis, is referred to as ``Burnup Credit.`` Burnup credit uses the actual physical composition of the fuel and accounts for the net reduction of fissile material and the buildup of neutron absorbers in the fuel as it is irradiated. Neutron absorbers include actinides and other isotopes generated as a result of the fission process. Using only the change in actinide isotopes in the burnup credit criticality analysis is referred to as ``Actinide-Only Burnup Credit.`` The use of burnup credit in the design of criticality control systems enables more spent fuel to be placed in a package. Increased package capacity results in a reduced number of storage, shipping and disposal containers for a given number of SNF assemblies. Fewer shipments result in a lower risk of accidents associated with the handling and transportation of spent fuel, thus reducing both radiological and nonradiological risk to the public. This paper describes the modeling and the results of comparison between measured and calculated isotopic inventories for a selected number of samples taken from a Yankee Rowe spent fuel assembly.

  17. Inherent safety of minimum-burnup breed and burn reactors

    SciTech Connect

    Qvist, S.; Reenspan, E.

    2012-07-01

    Reactors that aim to sustain the breed and burn (B and B) mode of operation at minimum discharge burnup require excellent neutron economy, Minimum-burnup B and B cores are generally large and feature low neutron leakage probability and a hard neutron spectrum. While highly promising fuel cycles can be achieved with such designs, the very same features are pushing the limits of the core's ability to passively respond safely to unprotected accidents. Low leakage minimum-burnup sodium-cooled B and B cores have a large positive coolant void-worth and coolant temperature reactivity coefficient. In this study, the applicability of major approaches for fast reactor void-worth reduction is evaluated specifically for B and B cores. The design, shuffling scheme and performance of a new metallic-fueled, sodium-cooled minimum burnup B and B core, used as basis for the void-worth reduction analysis, is presented. The analysis shows that reactivity control systems based on passive {sup 6}Li injection during temperature excursions are the only option able to provide negative void-worth without significantly increasing the minimum burnup required for sustaining the B and B mode of operation. A new type of lithium expansion module (LEM) system was developed specifically for B and B cores and its effect on core performance is presented. (authors)

  18. Local Burn-Up Effects in the NBSR Fuel Element

    SciTech Connect

    Brown N. R.; Hanson A.; Diamond, D.

    2013-01-31

    This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peaking relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.

  19. Effect of core polarizability on photoionization cross-section calculations.

    NASA Technical Reports Server (NTRS)

    Kirkpatrick, R. C.

    1972-01-01

    Demonstration of the importance of core polarizability in a case where cancellation is only moderate, with suggestion of an improvement to the scaled Thomas-Fermi (STF) wave functions of Stewart and Rotenberg (1965). The inclusion of dipole polarizability of the core for argon is shown to substantially improve the agreement between the theoretical and experimental photoionization cross sections for the ground-state configuration.

  20. A PWR Thorium Pin Cell Burnup Benchmark

    SciTech Connect

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  1. One method for calculating flux-MMF relationship of orthogonal-core

    SciTech Connect

    Tajima, Katsubumi; Kaga, Akio; Anazawa, Yoshihisa . Dept. of Electrical and Electronics Engineering); Ichinokura, Osamu . Dept. of Electrical Engineering)

    1993-11-01

    The orthogonal-core has various applications, e.g. as a variable inductor, a parametric transformer and a DC-AC converter. This paper describes one method for calculating the flux-MMF relationship of an orthogonal-core. The calculation is based on a 3-dimensional magnetic circuit model of the orthogonal-core. The model is derived by dividing the orthogonal-core, inclusive of the surrounding region, into elements comprising a 3-dimensional magnetic circuit. Using this model, the authors can compute the flux-MMF relationship of the orthogonal core with arbitrary dimensions from the B-H characteristic of the core material. The calculation method presented here is useful for optimum design of devices using an orthogonal-core.

  2. Displacement calculations across a metamorphic core complex mylonite zone: Pinaleno Mountains, southeastern Arizona

    SciTech Connect

    Naruk, S.J.

    1987-07-01

    Minimum offset of 7 km across the Pinaleno Mountains metamorphic core complex is calculated by integrating the shear strains across the exposed width of the mylonite zone. The calculated displacement equals the offset on the associated detachment fault, estimated from offset marker beds. The method of determining displacement by strain integration may be directly applicable to many other metamorphic core complexes.

  3. Validation Work to Support the Idaho National Engineering and Environmental Laboratory Calculational Burnup Methodology Using Shippingport Light Water Breeder Reactor (LWBR) Spent Fuel Assay Data

    SciTech Connect

    J. W. Sterbentz

    1999-08-01

    Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a depletion methodology previously employed to evaluate many of the radionuclide inventories for spent nuclear fuels at the Idaho National Engineering and Environmental Laboratory. The primary goal of the calculational task was to further support the validation of this particular calculational methodology and its application to diverse reactor types and fuels. Result comparisons between the calculated and measured mass concentrations in the three rods indicate good agreement for the three major uranium isotopes (U-233, U-234, U-235) with differences of less than 20%. For the seed and standard blanket rod, the U-233 and U-234 differences were within 5% of the measured values (these two isotopes alone represent greater than 97% of the EOL total uranium mass). For the major krypton and xenon fission product isotopes, differences of less than 20% and less than 30% were observed, respectively. In general, good agreement was obtained for nearly all the measured isotopes. For these isotopes exhibiting significant differences, possible explanations are discussed in terms of measurement uncertainty, complex transmutations, etc.

  4. Triton burnup measurements in KSTAR using a neutron activation system

    NASA Astrophysics Data System (ADS)

    Jo, Jungmin; Cheon, MunSeong; Kim, Jun Young; Rhee, T.; Kim, Junghee; Shi, Yue-Jiang; Isobe, M.; Ogawa, K.; Chung, Kyoung-Jae; Hwang, Y. S.

    2016-11-01

    Measurements of the time-integrated triton burnup for deuterium plasma in Korea Superconducting Tokamak Advanced Research (KSTAR) have been performed following the simultaneous detection of the d-d and d-t neutrons. The d-d neutrons were measured using a 3He proportional counter, fission chamber, and activated indium sample, whereas the d-t neutrons were detected using activated silicon and copper samples. The triton burnup ratio from KSTAR discharges is found to be in the range 0.01%-0.50% depending on the plasma conditions. The measured burnup ratio is compared with the prompt loss fraction of tritons calculated with the Lorentz orbit code and the classical slowing-down time. The burnup ratio is found to increase as plasma current and classical slowing-down time increase.

  5. /sup 18/O as a core plus two valence neutrons: A three-body Faddeev calculation

    SciTech Connect

    Ueta, K.; Miyake, H.; Mizukami, A.

    1983-01-01

    The nucleus /sup 18/O is studied assuming a three-body model: two neutrons outside an inert core of /sup 16/O: and solving the Faddeev equations. The calculated spectrum is in good agreement with experiment.

  6. Investigation of intravalence, core-valence and core-core electron correlation effects in polonium atomic structure calculations

    NASA Astrophysics Data System (ADS)

    Quinet, Pascal

    2014-09-01

    A detailed investigation of the atomic structure and radiative parameters involving the lowest states within the 6p4, 6p36d, 6p37s, 6p37p and 6p37d configurations of neutral polonium is reported in the present paper. Using different physical models based on the pseudo-relativistic Hartree-Fock approach, the influence of intravalence, core-valence and core-core electron correlation on the atomic parameters is discussed in detail. This work allowed us to fix the spectroscopic designation of some experimental level energy values and to provide for the first time a set of reliable oscillator strengths corresponding to 31 Po I spectral lines in the wavelength region from 175 to 987 nm.

  7. Monte-Carlo Continuous Energy Burnup Code System.

    2007-08-31

    Version 00 MCB is a Monte Carlo Continuous Energy Burnup Code for a general-purpose use to calculate a nuclide density time evolution with burnup or decay. It includes eigenvalue calculations of critical and subcritical systems as well as neutron transport calculations in fixed source mode or k-code mode to obtain reaction rates and energy deposition that are necessary for burnup calculations. The MCB-1C patch file and data packages as distributed by the NEADB are verymore » well organized and are being made available through RSICC as received. The RSICC package includes the MCB-1C patch and MCB data libraries. Installation of MCB requires MCNP4C source code and utility programs, which are not included in this MCB distribution. They were provided with the now obsolete CCC-700/MCNP-4C package.« less

  8. Designing Critical Experiments in Support of Full Burnup Credit

    SciTech Connect

    Mueller, Don; Roberts, Jeremy A

    2008-01-01

    Burnup credit is the process of accounting for the negative reactivity due to fuel burnup and generation of parasitic absorbers over fuel assembly lifetime. For years, the fresh fuel assumption was used as a simple bound in criticality work for used fuel storage and transportation. More recently, major actinides have been included [1]. However, even this yields a highly conservative estimate in criticality calculations. Because of the numerous economical benefits including all available negative reactivity (i.e., full burnup credit) could provide [2], it is advantageous to work toward full burnup credit. Unfortunately, comparatively little work has been done to include non-major actinides and other fission products (FP) in burnup credit analyses due in part to insufficient experimental data for validation of codes and nuclear data. The Burnup Credit Criticality Experiment (BUCCX) at Sandia National Laboratory was a set of experiments with {sup 103}Rh that have relevance for burnup credit [3]. This work uses TSUNAMI-3D to investigate and adjust a BUCCX model to match isotope-specific, energy-dependent k{sub eff} sensitivity profiles to those of a representative high-capacity cask model (GBC-32) [4] for each FP of interest. The isotopes considered are {sup 149}Sm, {sup 143}Nd, {sup 103}Rh, {sup 133}Cs, {sup 155}Gd, {sup 152}Sm, {sup 99}Tc, {sup 145}Nd, {sup 153}Eu, {sup 147}Sm, {sup 109}Ag, {sup 95}Mo, {sup 150}Sm, {sup 101}Ru, and {sup 151}Eu. The goal is to understand the biases and bias uncertainties inherent in nuclear data, and ultimately, to apply these in support of full burnup credit.

  9. Development of the MCNPX depletion capability: A Monte Carlo linked depletion method that automates the coupling between MCNPX and CINDER90 for high fidelity burnup calculations

    NASA Astrophysics Data System (ADS)

    Fensin, Michael Lorne

    Monte Carlo-linked depletion methods have gained recent interest due to the ability to more accurately model complex 3-dimesional geometries and better track the evolution of temporal nuclide inventory by simulating the actual physical process utilizing continuous energy coefficients. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a high-fidelity completely self-contained Monte-Carlo-linked depletion capability in a well established, widely accepted Monte Carlo radiation transport code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross section data permits in order to achieve a highly accurate temporal nuclide inventory solution. This work chronicles relevant nuclear history, surveys current methodologies of depletion theory, details the methodology in applied MCNPX and provides benchmark results for three independent OECD/NEA benchmarks. Relevant nuclear history, from the Oklo reactor two billion years ago to the current major United States nuclear fuel cycle development programs, is addressed in order to supply the motivation for the development of this technology. A survey of current reaction rate and temporal nuclide inventory techniques is then provided to offer justification for the depletion strategy applied within MCNPX. The MCNPX depletion strategy is then dissected and each code feature is detailed chronicling the methodology development from the original linking of MONTEBURNS and MCNP to the most recent public release of the integrated capability (MCNPX 2.6.F). Calculation results of the OECD/NEA Phase IB benchmark, H. B. Robinson benchmark and OECD/NEA Phase IVB are then provided. The acceptable results of these calculations offer sufficient confidence in the predictive capability of the MCNPX depletion method. This capability sets up a significant foundation, in a well established

  10. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    SciTech Connect

    Hartini, Entin Andiwijayakusuma, Dinan

    2014-09-30

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  11. Issues related to criticality safety analysis for burnup credit applications

    SciTech Connect

    DeHart, M.D.; Parks, C.V.

    1995-12-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh fuel loading assumption. Parametric analyses are required to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models are evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. This paper discusses the results of studies to determine the effect of two important modeling assumptions on the criticality analysis of pressurized-water reactor (PWR) spent fuel: (1) the effect of assumed burnup history (i.e., specific power during and time-dependent variations in operational power) during depletion calculations, and (2) the effect of axial burnup distributions on the neutron multiplication factor calculated for a three-dimensional (3-D) conceptual cask design.

  12. Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

    NASA Astrophysics Data System (ADS)

    Karim, Julia Abdul

    2008-05-01

    The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

  13. Burnup estimation of fuel sourcing radioactive material based on monitored Cs and Pu isotopic activity ratios in Fukushima N. P. S. accident

    SciTech Connect

    Yamamoto, T.; Suzuki, M.; Ando, Y.

    2012-07-01

    After the severe core damage of Fukushima Dai-Ichi Nuclear Power Station, radioactive material leaked from the reactor buildings. As part of monitoring of radioactivity in the site, measurements of radioactivity in soils at three fixed points have been performed for {sup 134}Cs and {sup 137}Cs with gamma-ray spectrometry and for Pu, Pu, and {sup 240}Pu with {alpha}-ray spectrometry. Correlations of radioactivity ratios of {sup 134}Cs to {sup 137}Cs, and {sup 238}Pu to the sum of {sup 239}Pu and {sup 240}Pu with fuel burnup were studied by using theoretical burnup calculations and measurements on isotopic inventories, and compared with the Cs and Pu radioactivity rations in the soils. The comparison indicated that the burnup of the fuel sourcing the radioactivity was from 18 to 38 GWd/t, which corresponded to that of the fuel in the highest power and, therefore, the highest decay heat in operating high-burnup fueled BWR cores. (authors)

  14. Investigation of the Fundamental Constants Stability Based on the Reactor Oklo Burn-Up Analysis

    NASA Astrophysics Data System (ADS)

    Onegin, M. S.; Yudkevich, M. S.; Gomin, E. A.

    2012-12-01

    The burn-up of few samples of the natural Oklo reactor zones 3, 5 was calculated using the modern Monte Carlo code. We reconstructed the neutron spectrum in the core by means of the isotope ratios: 147Sm/148Sm and 176Lu/175Lu. These ratios unambiguously determine the water content and core temperature. The isotope ratio of the 149Sm in the sample calculated using this spectrum was compared with experimental one. The disagreement between these two values allows one to limit a possible shift of the low lying resonance of 149Sm. Then, these limits were converted to the limits for the change of the fine structure constant α. We have found out, that for the rate of α change, the inequality ěrt˙ {α }/α ěrt<= 5× 10-18 is fulfilled, which is one order higher than our previous limit.

  15. Influence of porosity formation on irradiated UO2 fuel thermal conductivity at high burnup

    NASA Astrophysics Data System (ADS)

    Roostaii, B.; Kazeminejad, H.; Khakshournia, S.

    2016-10-01

    Based on the existing low temperature high burnup gaseous swelling model for UO2 fuel, the matrix swelling terms are calculated and the formation of total volume porosity up to burnup of 120 MWd/KgU is computed. The irradiated UO2 thermal conductivity model based on the Maxwell-Eucken correlation for porosity factor is selected as a case study and the calculation of porosity evolution with burnup is carried out. It is shown that taking into account the formation of porosity with burnup compared to the case with constant porosity equal to as-fabricated value leads to a decrease in the UO2 fuel thermal conductivity up to 15% at high burnup values of 120 MWd/kgU. Results of the calculations are also compared with the available experimental data and good agreement was found. The conducted parametric study clearly demonstrated the impact of the key parameters on the results of the present investigation.

  16. End effect K{sub eff} bias curve for actinide-only burnup credit casks

    SciTech Connect

    Kang, C.H.; Lancaster, D.B.

    1997-03-01

    A conservative end effect k{sub eff} bias curve for actinide-only burnup credit for spent fuel casks is presented in this paper. The k{sub eff} bias values can be added to the uniform axial burnup analysis to conservatively bound the actinide-only end effect. A normalized axial burnup distribution for the standard Westinghouse 17 x 17 assembly design is used for calculating k{sub eff}. The end effect calculated is a strong function of burnup, and increases as cask size size decreases. The presence of poison plates increases the end effect. The bias curve presented is based on the most limiting cask configuration of a single PWR assembly with completely black poison plates. Therefore, axially uniform criticality calculations with application of the proposed k{sub eff} could eliminate the need for axially burnup dependent analyses. 7 refs., 1 fig.

  17. Time delays between core power production and external detector response from Monte Carlo calculations

    SciTech Connect

    Valentine, T.E.; Mihalczo, J.T.

    1996-08-01

    One primary concern for design of safety systems for reactors is the time response of external detectors to changes in the core. This paper describes a way to estimate the time delay between the core power production and the external detector response using Monte Carlo calculations and suggests a technique to measure the time delay. The Monte Carlo code KENO-NR was used to determine the time delay between the core power production and the external detector response for a conceptual design of the Advanced Neutron Source (ANS) reactor. The Monte Carlo estimated time delay was determined to be about 10 ms for this conceptual design of the ANS reactor.

  18. Accounting for strong localized heterogeneities and local transport effect in core calculations

    SciTech Connect

    Ruggieri, J.M.; Doriath, J.Y.; Finck, P.J.; Boyer, R.

    1996-09-01

    Two methods based on the variational nodal transport method have been developed to account for localized heterogeneities and local transport effects in full core calculations. A local mesh refinement technique relies on using the projected partial ingoing surface currents produced during coarse-mesh iterations as boundary conditions for fine-mesh calculations embedded within the coarse-mesh calculations. The outgoing fine-mesh partial currents are averaged to serve in the coarse-mesh iterations. Then, a mixed transport-diffusion method using two levels of angular approximations for the surface partial currents depending on the node considered has been implemented to account for local transport effects in full core diffusion calculations. These methods have been tested for a model of the Superphenix complementary shutdown rods.

  19. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    SciTech Connect

    Ade, Brian J; Marshall, William BJ J; Martinez-Gonzalez, Jesus S

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  20. Core correlation effects in multiconfiguration calculations of isotope shifts in Mg i

    NASA Astrophysics Data System (ADS)

    Filippin, Livio; Godefroid, Michel; Ekman, Jörgen; Jönsson, Per

    2016-06-01

    The present work reports results from systematic multiconfiguration Dirac-Hartree-Fock calculations of isotope shifts for several well-known transitions in neutral magnesium. Relativistic normal and specific mass shift factors as well as the electronic probability density at the origin are calculated. Combining these electronic quantities with available nuclear data, energy and transition level shifts are determined for the 26Mg-24Mg pair of isotopes. Different models for electron correlation are adopted. It is shown that, although valence and core-valence models provide accurate values for the isotope shifts, the inclusion of core-core excitations in the computational strategy significantly improves the accuracy of the transition energies and normal mass shift factors.

  1. VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

    SciTech Connect

    Ellis, RJ

    2001-06-01

    The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.

  2. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    SciTech Connect

    Gauld, I.C.

    2002-12-23

    Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

  3. Benchmark calculation of no-core Monte Carlo shell model in light nuclei

    SciTech Connect

    Abe, T.; Shimizu, N.; Maris, P.; Vary, J. P.; Otsuka, T.; Utsuno, Y.

    2011-05-06

    The Monte Carlo shell model is firstly applied to the calculation of the no-core shell model in light nuclei. The results are compared with those of the full configuration interaction. The agreements between them are within a few % at most.

  4. Use of Relativistic Effective Core Potentials in the Calculation of Electron-Impact Ionization Cross Sections

    NASA Technical Reports Server (NTRS)

    Huo, Winifred M.; Kim, Yong-Ki

    1999-01-01

    Based on the Binary-Encounter-Bethe (BEB) model, the advantage of using relativistic effective core potentials (RECP) in the calculation of total ionization cross sections of heavy atoms or molecules containing heavy atoms is discussed. Numerical examples for Ar, Kr, Xe, and WF6 are presented.

  5. High-resolution Valence and Core Excitation Spectra via First-Principles Calculations and Experiment

    NASA Astrophysics Data System (ADS)

    Shirley, Eric; Fossard, F.; Gilmore, K.; Hug, G.; Kas, J. J.; Rehr, J. J.; Vila, F.

    We calculate the optical and C K-edge near edge spectra of crystalline and molecular C60 measured with high-resolution electron energy-loss spectroscopy. The calculations are carried out using at least three different methods: Bethe-Salpeter calculations using the NIST Bethe-Salpeter Equation solver (NBSE) in the valence and OCEAN (Obtaining Core Excitation with Ab initio methods and NBSE) suite [Gilmore et al., Comp. Phys. Comm., (2015)]; excited-core-hole calculations using XCH [D. Prendergast and G. Galli, Phys. Rev. Lett. 96, 215502 (2006)]; and constrained occupancy using StoBe (Stockholm-Berlin core-excitation code) [StoBe-deMon version 3.0, K. Hermann et al. (2009)]. They include self-energy effects, lifetime-damping, and Debye-Waller effects. A comparison of spectral features to those observed illustrates the sensitivity of certain features to computation details (e.g., self-energy corrections and core-hole screening). This may point to limitations of various approximations, e.g. in conventional BSE paradigm and/or the incomplete treatment of vibrational effects. Supported in part by DOE BES Grant DE-FG03-97ER45623 (JJR, JJK, FV).

  6. Criticality validation for burnup credit using recycle Pu criticals

    SciTech Connect

    Fuentes, E.; Lancaster, D.

    1997-04-01

    A set of 23 additional critical experiments were analyzed to provide additional input to the criticality validation portion of spent fuel cask analysis. The results of this analyses were combined with the previously analyzed criticals to determine the upper safety limit on k{sub eff}. The combined set of criticals can be used used for criticality validation for burnup credit, and are better suited for the range of isotopics in spent nuclear fuels. A trend observed in the analysis was that the calculated k{sub eff} deviates from the criticals in the positive direction, implying that increased burnup results in increased safety margin. 6 refs., 2 figs., 1 tab.

  7. Examination of temperature dependent subgroup formulations in direct whole core transport calculation for power reactors

    SciTech Connect

    Jung, Y. S.; Lee, U. C.; Joo, H. G.

    2012-07-01

    The traditional subgroup method which has been applied for lattice transport calculations has an inherent limitation for non-uniform temperature distributions. As a measure to incorporate temperature dependence into the subgroup formulation, the subgroup level and number density adjustment method have been proposed. In this paper, the temperature dependent subgroup formulations employed for reflecting the non-uniform temperature effects on the resonance spatial self-shielding are examined for the whole core transport calculation with the thermal feedback. For 2D pin-cell problem with non-uniform temperature profiles, the inherent limitation of conventional subgroup method is confirmed. And the improvement in terms of reactivity is observed with the proposed adjustment scheme. For the real PWR core calculation with thermal feedback in the hot-full-power condition, the noticeable correction for the fuel temperature coefficient by about 10% more negative is obtained with the correction schemes. (authors)

  8. High-Order Homogenization Method in Diffusion Theory for Reactor Core Simulation and Design Calculation

    SciTech Connect

    Farzad Rahnema

    2003-09-30

    Most modern nodal methods in use by the reactor vendors and utilities are based on the generalized equivalence theory (GET) that uses homogenized cross sections and flux discontinuity factors. These homogenized parameters, referred to as infinite medium parameters, are precomputed by performing single bundle fine-mesh calculations with zero current boundary conditions. It is known that for configurations in which the node-to-node leakage (e.g., surface current-to-flux ratio) is large the use of the infinite medium parameters could lead to large errors in the nodal solution. This would be the case for highly heterogeneous core configurations, typical of modern reactor core designs.

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

    SciTech Connect

    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.

  10. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    SciTech Connect

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

    2014-09-30

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided.

  11. VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4

    SciTech Connect

    Ellis, RJ

    2001-02-02

    The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.

  12. Liquid iron-sulfur alloys at outer core conditions by first-principles calculations

    NASA Astrophysics Data System (ADS)

    Umemoto, Koichiro; Hirose, Kei; Imada, Saori; Nakajima, Yoichi; Komabayashi, Tetsuya; Tsutsui, Satoshi; Baron, Alfred Q. R.

    2014-10-01

    We perform first-principles calculations to investigate liquid iron-sulfur alloys (Fe, Fe56S8, Fe52S12, and Fe48S16) under high-pressure and high-temperature (150-300 GPa and 4000-6000 K) conditions corresponding to the Earth's outer core. Considering only the density profile, the best match with the preliminary reference Earth model is by liquid Fe-14 wt % S (Fe50S14), assuming sulfur is the only light element. However, its bulk sound velocity is too high, in particular in the deep outer core, suggesting that another light component such as oxygen is required. An experimental check using inelastic X-ray scattering shows good agreement with the calculations. In addition, a present study demonstrates that the Birch's law does not hold for liquid iron-sulfur alloy, consistent with a previous report on pure liquid iron.

  13. Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd/t

    SciTech Connect

    Caruso, S.; Murphy, M.; Jatuff, F.; Chawla, R.

    2006-07-01

    High-resolution gamma spectroscopy has been employed for the measurement of {sup 134}Cs/{sup 137}Cs, {sup 154}Eu/{sup 137}Cs and {sup 134}Cs/{sup 154}Eu gamma intensity ratios from spent fuel with the purpose of deriving pin-averaged single-ratio burnup indicators for high and ultra-high burnups. Two UO{sub 2} pressurised water reactor (PWR) fuel rod segments with record burnup levels >80 GWd/t have been experimentally characterised. Additionally, pin cell depletion calculations have been performed for each sample with the deterministic code CASMO-4, using both its JEF2.2- and its ENDF/B-IV-based libraries, for three different descriptions of the fuel rod irradiation histories, in order to test the sensitivity of the results to neutron cross sections and to the depletion model employed. Measured and calculated ratios have then been compared. It is shown that the {sup 134}Cs/{sup 137}Cs ratio, frequently used as burnup monitor, is considerably less accurate for values exceeding 50 GWd/t; discrepancies of up to {approx}25% are found between measured and calculated values. The ratios built with the {sup 154}Eu concentration show much larger discrepancies, essentially because this isotope is rather poorly predicted as revealed by the use of different basic cross section data. (authors)

  14. THEHYCO-3DT: Thermal hydrodynamic code for the 3 dimensional transient calculation of advanced LMFBR core

    SciTech Connect

    Vitruk, S.G.; Korsun, A.S.; Ushakov, P.A.

    1995-09-01

    The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors.

  15. Methodology of full-core Monte Carlo calculations with leakage parameter evaluations for benchmark critical experiment analysis

    NASA Astrophysics Data System (ADS)

    Sboev, A. G.; Ilyashenko, A. S.; Vetrova, O. A.

    1997-02-01

    The method of bucking evaluation, realized in the MOnte Carlo code MCS, is described. This method was applied for calculational analysis of well known light water experiments TRX-1 and TRX-2. The analysis of this comparison shows, that there is no coincidence between Monte Carlo calculations, obtained by different ways: the MCS calculations with given experimental bucklings; the MCS calculations with given bucklings evaluated on base of full core MCS direct simulations; the full core MCNP and MCS direct simulations; the MCNP and MCS calculations, where the results of cell calculations are corrected by the coefficients taking into the account the leakage from the core. Also the buckling values evaluated by full core MCS calculations have differed from experimental ones, especially in the case of TRX-1, when this difference has corresponded to 0.5 percent increase of Keff value.

  16. Gross Gamma Dose Rate Measurements for TRIGA Spent Nuclear Fuel Burnup Validation

    SciTech Connect

    Winston, Philip Lon; Sterbentz, James William

    2001-04-01

    Gross gamma-ray dose rates from six spent TRIGA fuel elements were measured and compared to calculated values as a means to validate the reported element burnups. A newly installed and functional gamma-ray detection subsystem of the In-Cell Examination System was used to perform the measurements and is described in some detail. The analytical methodology used to calculate the corresponding dose rates is presented along with the calculated values. Comparison of the measured and calculated dose rates for the TRIGA fuel elements indicates good agreement (less than a factor of 2 difference). The intent of the subsystem is to measure the gross gamma dose rate and correlate the measurement to a calculated dose rate based on the element s known burnup and other pertinent spent fuel information. Although validation of the TRIGA elements’ burnup is of primary concern in this paper, the measurement and calculational techniques can be used to either validate an element’s reported burnup or provide a burnup estimate for an element with an unknown burnup.

  17. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR

    SciTech Connect

    Hanson, A.L.; Diamond, D.

    2011-09-30

    A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Previously, the design of the LEU fuel had been determined in order to provide the users of the NBSR with the same cycle length as exists for the current HEU fueled reactor. The fuel composition at different points within an equilibrium fuel cycle had also been determined. In the present study, neutronics parameters have been calculated for these times in the fuel cycle for both the existing HEU and the proposed LEU equilibrium cores. The results showed differences between the HEU and LEU cores that would not lead to any significant changes in the safety analysis for the converted core. In general the changes were reasonable except that the figure-of-merit for neutrons that can be used by experimentalists shows there will be a 10% reduction in performance. The calculations included kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions.

  18. Use of Th and U in CANDU-6 and ACR-700 on the once-through cycle: Burnup analyses, natural U requirement/saving and nuclear resource utilization

    NASA Astrophysics Data System (ADS)

    Türkmen, Mehmet; Zabunoğlu, Okan H.

    2012-10-01

    Use of U and U-Th fuels in CANDU type of reactors (CANDU-6 and ACR-700) on the once-through nuclear fuel cycle is investigated. Based on the unit-cell approximation with the homogeneous-bundle/core model, utilizing the MONTEBURNS code, burnup computations are performed; discharge burnups are determined and expressed as functions of 235U and Th fractions, when applicable. Natural Uranium Requirement (and Saving) and Nuclear Resource Utilization are calculated for varying fuel compositions. Results are analyzed to observe the effects of 235U and Th fractions, thus to reach conclusions about use of Th in CANDU-6 and ACR-700 on the once-through cycle.

  19. Liquid iron-hydrogen alloys at outer core conditions by first-principles calculations

    NASA Astrophysics Data System (ADS)

    Umemoto, Koichiro; Hirose, Kei

    2015-09-01

    We examined the density, bulk sound (compressional) velocity, and Grüneisen parameter of liquid pure Fe, Fe100H28 (0.50 wt % H), Fe88H40 (0.81 wt % H), and Fe76H52 (1.22 wt % H) at Earth's outer core pressure and temperature (P-T) conditions (~100 to 350 GPa, 4000 to 7000 K) based on first-principles molecular dynamics calculations. The results demonstrate that the thermodynamic Grüneisen parameter of liquid iron alloy decreases with increasing pressure, temperature, and hydrogen concentration, indicating a relatively small temperature gradient in the outer core when hydrogen is present. Along such temperature profile, both the density and compressional velocity of liquid iron containing ~1 wt % hydrogen match seismological observations. It suggests that hydrogen could be a primary light element in the core, although the shear velocity of the inner core is not reconciled with solid Fe-H alloy and thus requires another impurity element.

  20. Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

    SciTech Connect

    Wagner, John C; Parks, Cecil V; Mueller, Don; Gauld, Ian C

    2010-01-01

    Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and

  1. Phenomena and Parameters Important to Burnup Credit

    SciTech Connect

    Parks, C.V.

    2001-01-10

    Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given.

  2. Overview of spent fuel burnup measurements

    SciTech Connect

    McLeod, N.B. Associates, Inc., Oakton, VA )

    1991-01-01

    The purpose of this paper is to provide a summary of the principal factors that influence burnup measurement accuracy, and the manner in which different combinations of these factors in any particular application influence the achievable accuracy in that application. The principal focus of the paper is on the passive measurement of various burnup indicators. This paper also provides a general background for four subsequent papers which discuss burnup measurement in two particular applications: for burnup credit in spent fuel shipping casks; and for safeguards purposes. This paper provides a basis for comparing these two applications in terms that directly relate to the measurement process.

  3. Radial Power Profile of MOX and LEU Fuel Pellet Versus Burnup

    SciTech Connect

    Chang, Gray S.; Pedersen, Robert C.

    2002-07-01

    One of challenge to burn the WG-Pu in Mixed Oxide (MOX) fuel in light water reactors (LWR) is to demonstrate that the differences between WG-MOX, RG-MOX, and LWR LEU fuel are minimal, and therefore, the commercial MOX and LEU fuel experience base is applicable. The MCWO-calculated Radial Power Profile of LEU, Weapons Grade-MOX and Reactor Grade-MOX fuel pellets at various burnups are similar toward the end of life (50 GWd/t). Therefore, the LEU fuel performance evaluation code - FRAPCON-3 with modifications, such as, the detailed fission power profiles versus burnup, can be used in the MOX fuel pellet performance analysis. MCWO also calculated the {sup 240}Pu/Pu ratio in WG-MOX versus burnup, which reaches an average of 31.25% at discharged burnup of 50 GWd/t. It meets the spent fuel standard for WG-Pu disposition in LWR. (authors)

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

    SciTech Connect

    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.

  5. Energy-Deposition and Damage Calculations in Core-Vessel Inserts at the Spallation Neutron Source

    SciTech Connect

    Murphy, B.D.

    2002-06-25

    Heat-deposition and damage calculations are described for core-vessel inserts in the target area of the Spallation Neutron Source. Two separate designs for these inserts (or neutron beam tubes) were studied; a single-unit insert and a multi-unit insert. The single unit contains a neutron guide; the multi unit does not. Both units are constructed of stainless steel. For the single unit, separate studies were carried out with the guide composed of stainless steel, glass, and aluminum. Results are also reported for an aluminum window on the front of the insert, a layer of nickel on the guide, a cadmium shield surrounding the guide, and a stainless steel plug in the beam-tube opening. The locations of both inserts were the most forward positions to be occupied by each design respectively thus ensuring that the calculations are conservative.

  6. Hybrid parallel code acceleration methods in full-core reactor physics calculations

    SciTech Connect

    Courau, T.; Plagne, L.; Ponicot, A.; Sjoden, G.

    2012-07-01

    When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

  7. Calculated coupling efficiency between an elliptical-core optical fiber and an optical waveguide over temperature

    NASA Technical Reports Server (NTRS)

    Tuma, Margaret L.; Weisshaar, Andreas; Li, Jian; Beheim, Glenn

    1995-01-01

    To determine the feasibility of coupling the output of a single-mode optical fiber into a single-mode rib waveguide in a temperature varying environment, a theoretical calculation of the coupling efficiency between the two was investigated. Due to the complex geometry of the rib guide, there is no analytical solution to the wave equation for the guided modes, thus, approximation and/or numerical techniques must be utilized to determine the field patterns of the guide. In this study, three solution methods were used for both the fiber and guide fields; the effective-index method (EIM), Marcatili's approximation, and a Fourier method. These methods were utilized independently to calculate the electric field profile of each component at two temperatures, 20 C and 300 C, representing a nominal and high temperature. Using the electric field profile calculated from each method, the theoretical coupling efficiency between an elliptical-core optical fiber and a rib waveguide was calculated using the overlap integral and the results were compared. It was determined that a high coupling efficiency can be achieved when the two components are aligned. The coupling efficiency was more sensitive to alignment offsets in the y direction than the x, due to the elliptical modal field profile of both components. Changes in the coupling efficiency over temperature were found to be minimal.

  8. High accuracy modeling for advanced nuclear reactor core designs using Monte Carlo based coupled calculations

    NASA Astrophysics Data System (ADS)

    Espel, Federico Puente

    The main objective of this PhD research is to develop a high accuracy modeling tool using a Monte Carlo based coupled system. The presented research comprises the development of models to include the thermal-hydraulic feedback to the Monte Carlo method and speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Presently, deterministic codes based on the diffusion approximation of the Boltzmann transport equation, coupled with channel-based (or sub-channel based) thermal-hydraulic codes, carry out the three-dimensional (3-D) reactor core calculations of the Light Water Reactors (LWRs). These deterministic codes utilize nuclear homogenized data (normally over large spatial zones, consisting of fuel assembly or parts of fuel assembly, and in the best case, over small spatial zones, consisting of pin cell), which is functionalized in terms of thermal-hydraulic feedback parameters (in the form of off-line pre-generated cross-section libraries). High accuracy modeling is required for advanced nuclear reactor core designs that present increased geometry complexity and material heterogeneity. Such high-fidelity methods take advantage of the recent progress in computation technology and coupled neutron transport solutions with thermal-hydraulic feedback models on pin or even on sub-pin level (in terms of spatial scale). The continuous energy Monte Carlo method is well suited for solving such core environments with the detailed representation of the complicated 3-D problem. The major advantages of the Monte Carlo method over the deterministic methods are the continuous energy treatment and the exact 3-D geometry modeling. However, the Monte Carlo method involves vast computational time. The interest in Monte Carlo methods has increased thanks to the improvements of the capabilities of high performance computers. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods

  9. Instant release fraction and matrix release of high burn-up UO2 spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    NASA Astrophysics Data System (ADS)

    Serrano-Purroy, D.; Clarens, F.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; de Pablo, J.; Casas, I.; Giménez, J.; Martínez-Esparza, A.

    2012-08-01

    Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and

  10. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    SciTech Connect

    Salay, Michael; Gauntt, Randall O.; Lee, Richard Y.; Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  11. Properties of metastable alkaline-earth-metal atoms calculated using an accurate effective core potential

    SciTech Connect

    Santra, Robin; Christ, Kevin V.; Greene, Chris H.

    2004-04-01

    The first three electronically excited states in the alkaline-earth-metal atoms magnesium, calcium, and strontium comprise the (nsnp){sup 3}P{sub J}{sup o}(J=0,1,2) fine-structure manifold. All three states are metastable and are of interest for optical atomic clocks as well as for cold-collision physics. An efficient technique--based on a physically motivated potential that models the presence of the ionic core--is employed to solve the Schroedinger equation for the two-electron valence shell. In this way, radiative lifetimes, laser-induced clock shifts, and long-range interaction parameters are calculated for metastable Mg, Ca, and Sr.

  12. Ab Initio No-Core Shell Model Calculations Using Realistic Two- and Three-Body Interactions

    SciTech Connect

    Navratil, P; Ormand, W E; Forssen, C; Caurier, E

    2004-11-30

    There has been significant progress in the ab initio approaches to the structure of light nuclei. One such method is the ab initio no-core shell model (NCSM). Starting from realistic two- and three-nucleon interactions this method can predict low-lying levels in p-shell nuclei. In this contribution, we present a brief overview of the NCSM with examples of recent applications. We highlight our study of the parity inversion in {sup 11}Be, for which calculations were performed in basis spaces up to 9{Dirac_h}{Omega} (dimensions reaching 7 x 10{sup 8}). We also present our latest results for the p-shell nuclei using the Tucson-Melbourne TM three-nucleon interaction with several proposed parameter sets.

  13. Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

    NASA Astrophysics Data System (ADS)

    Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier

    2014-06-01

    As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard

  14. ATR WG-MOX Fuel Pellet Burnup Measurement by Monte Carlo - Mass Spectrometric Method

    SciTech Connect

    Chang, Gray Sen I

    2002-10-01

    This paper presents a new method for calculating the burnup of nuclear reactor fuel, the MCWO-MS method, and describes its application to an experiment currently in progress to assess the suitability for use in light-water reactors of Mixed-OXide (MOX) fuel that contains plutonium derived from excess nuclear weapons material. To demonstrate that the available experience base with Reactor-Grade Mixed uranium-plutonium OXide (RGMOX) can be applied to Weapons-Grade (WG)-MOX in light water reactors, and to support potential licensing of MOX fuel made from weapons-grade plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory. Fuel burnup is an important parameter needed for fuel performance evaluation. For the irradiated MOX fuel’s Post-Irradiation Examination, the 148Nd method is used to measure the burnup. The fission product 148Nd is an ideal burnup indicator, when appropriate correction factors are applied. In the ATR test environment, the spectrum-dependent and burnup-dependent correction factors (see Section 5 for detailed discussion) can be substantial in high fuel burnup. The validated Monte Carlo depletion tool (MCWO) used in this study can provide a burnup-dependent correction factor for the reactor parameters, such as capture-to-fission ratios, isotopic concentrations and compositions, fission power, and spectrum in a straightforward fashion. Furthermore, the correlation curve generated by MCWO can be coupled with the 239Pu/Pu ratio measured by a Mass Spectrometer (in the new MCWO-MS method) to obtain a best-estimate MOX fuel burnup. A Monte Carlo - MCWO method can eliminate the generation of few-group cross sections. The MCWO depletion tool can analyze the detailed spatial and spectral self-shielding effects in UO2, WG-MOX, and reactor-grade mixed oxide (RG-MOX) fuel pins. The MCWO-MS tool only

  15. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    SciTech Connect

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  16. ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation

    SciTech Connect

    G.S. Chang; P. A. Roth; M. A. Lillo

    2009-11-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is being studied to determine the feasibility of converting it from the highly enriched Uranium (HEU) fuel that is currently uses to low enriched Uranium (LEU) fuel. In order to achieve this goal, it would be best to qualify some different computational methods than those that have been used at ATR for the past 40 years. This paper discusses two methods of calculating the burnup of ATR fuel elements. The existing method, that uses the PDQ code, is compared to a modern method that uses A General Monte Carlo N-Particle Transport Code (MCNP) combined with the Origen2.2 code. This modern method, MCNP with ORIGEN2.2 (MCWO), is found to give excellent agreement with the existing method (PDQ). Both of MCWO and PDQ are also in a very good agreement to the 235U burnup data generated by an analytical method.

  17. Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

    SciTech Connect

    Jung, Y. S.; Joo, H. G.; Yoon, J. I.

    2013-07-01

    The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

  18. Calculated Actinide and Fission Product Concentration Ratios for Gaseous Effluent Monitoring Using Monteburns 3.01

    SciTech Connect

    Charlton, William S.; Perry, Robert T.; Fearey, Bryan L.; Parish, Theodore A.

    2000-08-15

    Techniques have been developed at Los Alamos National Laboratory for accurately calculating certain spent-fuel isotope concentration ratios for pressurized water reactor assemblies using a linked MCNP/ORIGEN2 code named Monteburns 3.01, without resorting to an assembly or full-core calculation. The effects of various fuel parameters such as the number of radial fuel regions per pin, burnup step size, reactor power, reactivity control mechanisms, and axial profiles have been studied. The significance of each factor was determined. A method was also proposed for calculating spent-fuel inventories as a function of burnup for a wide range of reactors and fuel types. It was determined that accurate calculations can be obtained using a three-dimensional, modified pin cell with seven radial fuel regions and two (flat-flux) axial fuel regions calculated with 2000 MWd/tonne U burnup steps for burnups ranging from 0 to 50 000 MWd/tonne U. The calculational technique was benchmarked to measured values from the Calvert Cliffs Unit 1 reactor, and good agreement from the point of view of calibrating a monitoring instrument was found for most cases.

  19. FaCE: a tool for three body Faddeev calculations with core excitation

    NASA Astrophysics Data System (ADS)

    Thompson, I. J.; Nunes, F. M.; Danilin, B. V.

    2004-08-01

    FaCE is a self contained program, with namelist input, that solves the three body Faddeev equations. It enables the inclusion of excitation of one of the three bodies, whilst the other two remain inert. It is particularly useful for obtaining the binding energies and bound state structure compositions of light exotic nuclei treated as three-body systems, given the three effective two body interactions. A large variety of forms for these interactions may be defined, and supersymmetric transformations of these potentials may be calculated whenever two body states need to be removed due to Pauli blocking. Program summaryTitle of program: FaCE (Faddeev with Core Excitation) Catalogue identifier: ADTW Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADTW Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Computers: The code is designed to run on any Unix/Linux workstation or PC. Operating systems: Linux or UNIX Program language used: Fortran-90 Numerical libraries used: Source code for 6 routines from the NAG and BLAS libraries is included to enable independent compilation. Memory required to execute with typical data: 9 Mbytes of RAM memory and 12 MB of hard disk space. No. of bits in a word: 32 or 64 No. of bytes in distributed program, including test data, etc.: 116 514 No. of lines in distributed program, including test data, etc.: 15 574 Distribution format: tar gzip file Nature of physical problem: The program calculates eigenenergies and eigenstates for the three body problem by solving the Faddeev equations. Method of solution: Given the two body effective potentials it performs the supersymmetric transformation in case where there are forbidden states to be removed. The three body wavefunction is expanded in hyperspherical coordinates, the hyper-angular part is a series of Jacobi polynomials and the hyper-radial part is written in terms of a Laguerre basis. Within this basis the three body matrix elements are calculated

  20. MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis

    SciTech Connect

    Gray S Chang

    2005-04-01

    The UNIX BASH (Bourne Again Shell) script MCWO has been developed at the Idaho National Engineering and Environment Laboratory (INEEL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN2. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. MCWO can handle a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) powers, and irradiation time intervals. The program processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN2, and data process module calculations are then output successively as the code runs. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion. The basic requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with ORIGEN2 and other calculations are performed by UNIX BASH script MCWO. This paper presents the MCWO-calculated results of the RERTR-1 and -2 , and the Weapons-Grade Mixed Oxiide fuel (Wg-MOX) fuel experiments in ATR and compares the MCWO-calculated results with the measured data.

  1. Development of burnup dependent fuel rod model in COBRA-TF

    NASA Astrophysics Data System (ADS)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  2. Lattice Cell Calculations, Slowing Down Theory and Computer Code Wims; Vver Type Reactors

    NASA Astrophysics Data System (ADS)

    Moen, J.; Brekke, A.; Hall, C.

    1991-01-01

    The following sections are included: * INTRODUCTION * WIMS AS A TOOL FOR REACTOR CORE CALCULATIONS * GENERAL STRUCTURE OF THE WIMS CODE * WIMS APPROACH TO THE SLOWING DOWN CALCULATIONS * MULTIGROUP OSCOPIC CROSS SECTIONS, RESONANCE TREATMENT * DETERMINATION OF MULTIGROUP SPECTRA * PHYSICAL MODELS IN MAIN TRANSPORT CALCULATIONS * BURNUP CALCULATIONS * APPLICATION OF WIMSD-4 TO VVER TYPE LATTICES * FINAL REMARKS * REFERENCES * APPENDIX A: DANCOFF FACTOR - STANDARD APPROACH * APPENDIX B: FORMULAS FOR DANCOFF AND BELL FACTORS CALCULATIONS APPLIED IN PREWIM * APPENDIX C: CALCULATION OF ONE GROUP PROBABILITIES Pij IN AN ANNULAR SYSTEM * APPENDIX D: SCHAEFER'S METHOD

  3. Extended Burnup Demonstration Reactor Fuels Program. Annual progress report, April 1983-March 1984. [BWR

    SciTech Connect

    Exarhos, C.A.

    1985-06-20

    The US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities Nuclear Corporation have participated since 1979 in a cooperative Extended Burnup Demonstration Program. Under the program, standard ENC-fabricated reload fuel in the Big Rock Point and Oyster Creek reactor cores has been irradiated to discharge burnups at or beyond 35,000 MWD/MTU, one to two cycles beyond its originally projected exposure life. The program provides for examination of the fuel at poolside before and after each extended burnup cycle as well as for limited destructive hot cell examination. The 1984 progress report covers work performed under the EBD program between April 1983 and March 1984. Major milestones reached during the period include completion of a hot cell examination on four high burnup rods from Big Rock Point and of a poolside on the Oyster Creek EBD fuel at discharge. The hot cell examination of four rods at burnups to 37.2 GWD/MTU confirmed poolside measurements on the same fuel, showing the urania and gadolinia-bearing fuel rods to be in excellent condition. No major cladding degradation, pellet restructuring, or pellet-clad interaction was found in any of the samples examined. The Oyster Creek fuel, examined at an assembly average exposure of 34.5 GWD/MTU, showed good performance with regard to both diametral creepdown and clad oxide accumulation.

  4. Calculation of the elastic properties of a triangular cell core for lightweight composite mirrors

    NASA Astrophysics Data System (ADS)

    Penado, F. Ernesto; Clark, James H., III; Walton, Joshua P.; Romeo, Robert C.; Martin, Robert N.

    2007-09-01

    The use of composite materials in the fabrication of optical telescope mirrors offers many advantages over conventional methods, including lightweight, portability and the potential for lower manufacturing costs. In the construction of the substrate for these mirrors, sandwich construction offers the advantage of even lower weight and higher stiffness. Generally, an aluminum or Nomex honeycomb core is used in composite applications requiring sandwich construction. However, the use of a composite core offers the potential for increased stiffness and strength, low thermal distortion compatible with that of the facesheets, the absence of galvanic corrosion and the ability to readily modify the core properties. In order to design, analyze and optimize these mirrors, knowledge of the mechanical properties of the core is essential. In this paper, the mechanical properties of a composite triangular cell core (often referred to as isogrid) are determined using finite element analysis of a representative unit cell. The core studied offers many advantages over conventional cores including increased thermal and dimensional stability, as well as low weight. Results are provided for the engineering elastic moduli of cores made of high stiffness composite material as a function of the ply layup and cell size. Finally, in order to illustrate the use of these properties in a typical application, a 1.4-m diameter composite mirror is analyzed using the finite element method, and the resulting stiffness and natural frequencies are presented.

  5. Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel

    SciTech Connect

    Wagner, J.C.

    2000-10-12

    /or decreasing the neutron absorber concentration. However, regulations associated with permanent disposal require consideration of scenarios and/or package conditions that are not relevant or credible for storage or transportation, and as a result, necessitate credit for burnup in BWR fuel to maintain capacity objectives. Burnup credit relies on depletion calculations to provide a conservative estimate of spent fuel contents and subsequent criticality calculations to assess the value of k{sub eff} for a spent fuel cask or a fuel configuration under a variety of postulated conditions. Therefore, validation is necessary to quantify biases and uncertainties between analytic predictions and measured isotopics. However, the design and operational aspects of BWRs result in a more heterogeneous and time-varying reactor configuration than those of PWRs. Thus, BWR spent fuel analyses and validation efforts are significantly more complicated than those of their PWR counterparts. BWR spent fuel assemblies are manufactured with variable enrichments, both radially and axially, are exposed to time- and spatially-varying void distributions, contain integral burnable absorber rods, and are subject to partial control-blade insertion during operation. The latter is especially true in older fuel assemblies. Away-from-reactor depletion tools used for characterization of spent fuel have typically been developed and validated for more homogeneous PWR fuel assemblies without integral burnable absorber rods, and thus must be reassessed for BWR configurations to determine a conservative methodology for estimating the isotopic content of spent BWR fuel. This report examines the use of SAS2H8 for calculating spent BWR fuel isotopics for burnup-credit criticality safety analyses and assesses the adequacy of SAS2H for this task. The effects of SAS2H modeling assumptions on calculated spent BWR fuel isotopics and the effects of depletion assumptions on calculated k{sub inf} values are investigated. Detailed

  6. Burnup Credit Approach Used in the Yucca Mountain License Application

    SciTech Connect

    Scaglione, John M; Wagner, John C

    2010-01-01

    The United States Department of Energy has submitted a license application (LA) for construction authorization of a deep geologic repository at Yucca Mountain, Nevada. The license application is currently under review by the United States Nuclear Regulatory Commission (NRC). This paper will describe the methodology and approach used in the LA to address the issue of criticality and the role of burnup credit during the postclosure period. The most significant and effective measures for prevention of criticality in the repository include multiple redundant barriers that act to isolate fissionable material from water (which can act as a moderator, corrosive agent, and transporter of fissile material); inherent geometry of waste package internals and waste forms; presence of fixed neutron absorbers in waste package internals; and fuel burnup for commercial spent nuclear fuel. A probabilistic approach has been used to screen criticality from the total system performance assessment. Within the probabilistic approach, criticality is considered an event, and the total probability of a criticality event occurring within 10,000 years of disposal is calculated and compared against the regulatory criterion. The total probability of criticality includes contributions associated with both internal (within waste packages) and external (external to waste packages) criticality for each of the initiating events that could lead to waste package breach. The occurrence of and conditions necessary for criticality in the repository have been thoroughly evaluated using a comprehensive range of parameter distributions. A simplified design-basis modeling approach has been used to evaluate the probability of criticality by using numerous significant and conservative assumptions. Burnup credit is used only for evaluations of in-package configurations and uses a combination of conservative and bounding modeling approximations to ensure conservatism. This paper will review the NRC regulatory

  7. Summary of high burnup fuel issues and NRC`s plan of action

    SciTech Connect

    Meyer, R.O.

    1997-01-01

    For the past two years the Office of Nuclear Regulatory Research has concentrated mostly on the so-called reactivity-initiated accidents -- the RIAs -- in this session of the Water Reactor Safety Information Meeting, but this year there is a more varied agenda. RIAs are, of course, not the only events of interest for reactor safety that are affected by extended burnup operation. Their has now been enough time to consider a range of technical issues that arise at high burnup, and a list of such issues being addressed in their research program is given here. (1) High burnup capability of the steady-state code (FRAPCON) used for licensing audit calculations. (2) General capability (including high burnup) of the transient code (FRAPTRAN) used for special studies. (3) Adequacy at high burnup of fuel damage criteria used in regulation for reactivity accidents. (4) Adequacy at high burnup of models and fuel related criteria used in regulation for loss-of-coolant accidents (LOCAs). (5) Effect of high burnup on fuel system damage during normal operation, including control rod insertion problems. A distinction is made between technical issues, which may or may not have direct licensing impacts, and licensing issues. The RIAs became a licensing issue when the French test in CABRI showed that cladding failures could occur at fuel enthalpies much lower than a value currently used in licensing. Fuel assembly distortion became a licensing issue when control rod insertion was affected in some operating plants. In this presentation, these technical issues will be described and the NRC`s plan of action to address them will be discussed.

  8. Accuracy of Effective Core Potentials and Basis Sets for Density Functional Calculations, Including Relativistic Effects, As Illustrated by Calculations on Arsenic Compounds.

    PubMed

    Xu, Xuefei; Truhlar, Donald G

    2011-09-13

    For molecules containing the fourth-period element arsenic, we test (i, ii) the accuracy of all-electron (AE) basis sets from the def2-xZVP and ma-xZVP series (where xZ is S, TZ, or QZ), (iii) the accuracy of the 6-311G series of AE basis sets with additional polarization and diffuse functions, and (iv) the performance of effective core potentials (ECPs). The first set of tests involves basis-set convergence studies with eleven density functionals for five cases: equilibrium dissociation energy (De) of As2, vertical ionization potential (VIP) of As2, IP of As, acid dissociation of H3AsO4, and De of FeAs. A second set of tests involves the same kinds of basis-set convergence studies for the VIP and De values of As3 and As4 clusters. Both relativistic and nonrelativistic calculations are considered, including in each case both AE calculations and calculations with ECPs. Convergence and accuracy are assessed by comparing to relativistic AE calculations with the cc-pV5Z-DK or ma-cc-pV5Z-DK basis and to nonrelativistic AE calculations with the cc-pV5Z or ma-cc-pV5Z basis. The primary objective of this study is to evaluate the abilities of ECPs with both their recommended basis sets and other basis sets to reproduce the results of all-electron relativistic calculations. The performance of the def2 and ma series basis sets is consistent with their sizes, and quadruple-ζ basis sets are the best. The def2-TZVP basis set performs better than most of the 6-311G series basis sets, which are the most commonly used basis sets in the previous studies of arsenic compounds. However, relativistic def2-TZVP calculations are not recommended. The large-core ECPs, which are the only available ECPs for arsenic in the popular Gaussian program, have average errors of 9-12 kcal/mol for the arsenic systems studied; therefore, these ECPs are not recommended. The triple-ζ small-core relativistic ECP (RECP) basis set cc-pVTZ-PP is found to have performance better than that of the def2-TZVP

  9. Verification study of thorium cross section in MVP calculation of thorium based fuel core using experimental data

    SciTech Connect

    Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T.; Takaki, N.; Yamaguchi, A.; Watanabe, H.; Unesaki, H.

    2012-07-01

    Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)

  10. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    SciTech Connect

    A.H. Wells

    2004-11-17

    The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).

  11. Burnup verification tests with the FORK measurement system-implementation for burnup credit

    SciTech Connect

    Ewing, R.I.

    1994-08-01

    Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. It was designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program and is well suited to verify burnup and cooling time records at commercial Pressurized Water Reactor (PWR) sites. This report deals with the application of the FORK system to burnup credit operations.

  12. Depletion analysis of the UMLRR reactor core using MCNP6

    NASA Astrophysics Data System (ADS)

    Odera, Dim Udochukwu

    Accurate knowledge of the neutron flux and temporal nuclide inventory in reactor physics calculations is necessary for a variety of application in nuclear engineering such as criticality safety, safeguards, and spent fuel storage. The Monte Carlo N- Particle (MCNP6) code with integrated buildup depletion code (CINDER90) provides a high-fidelity tool that can be used to perform 3D, full core simulation to evaluate fissile material utilization, and nuclide inventory calculations as a function of burnup. The University of Massachusetts Lowell Research Reactor (UMLRR) reactor has been modeled with the deterministic based code, VENTURE and with an older version of MCNP (MCNP5). The MIT developed MCODE (MCNP ORIGEN DEPLETION CODE) was used previously to perform some limited depletion calculations. This work chronicles the use of MCNP6, released in June 2013, to perform coupled neutronics and depletion calculation. The results are compared to previously benchmarked results. Furthermore, the code is used to determine the ratio of fission products 134Cs and 137Cs (burnup indicators), and the resultant ratio is compared to the burnup of the UMLRR.

  13. Model for evolution of grain size in the rim region of high burnup UO2 fuel

    NASA Astrophysics Data System (ADS)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

    The restructuring process of the high burnup structure (HBS) formation in UO2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results.

  14. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    SciTech Connect

    Ade, Brian J; Marshall, William BJ J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Martinez-Gonzalez, Jesus S

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  15. Sensitivity and Uncertainty Analysis to Burnup Estimates on ADS using the ACAB Code

    NASA Astrophysics Data System (ADS)

    Cabellos, O.; Sanz, J.; Rodríguez, A.; González, E.; Embid, M.; Alvarez, F.; Reyes, S.

    2005-05-01

    Within the scope of the Accelerator Driven System (ADS) concept for nuclear waste management applications, the burnup uncertainty estimates due to uncertainty in the activation cross sections (XSs) are important regarding both the safety and the efficiency of the waste burning process. We have applied both sensitivity analysis and Monte Carlo methodology to actinides burnup calculations in a lead-bismuth cooled subcritical ADS. The sensitivity analysis is used to identify the reaction XSs and the dominant chains that contribute most significantly to the uncertainty. The Monte Carlo methodology gives the burnup uncertainty estimates due to the synergetic/global effect of the complete set of XS uncertainties. These uncertainty estimates are valuable to assess the need of any experimental or systematic re-evaluation of some uncertainty XSs for ADS.

  16. Sensitivity and Uncertainty Analysis to Burn-up Estimates on ADS Using ACAB Code

    SciTech Connect

    Cabellos, O; Sanz, J; Rodriguez, A; Gonzalez, E; Embid, M; Alvarez, F; Reyes, S

    2005-02-11

    Within the scope of the Accelerator Driven System (ADS) concept for nuclear waste management applications, the burnup uncertainty estimates due to uncertainty in the activation cross sections (XSs) are important regarding both the safety and the efficiency of the waste burning process. We have applied both sensitivity analysis and Monte Carlo methodology to actinides burnup calculations in a lead-bismuth cooled subcritical ADS. The sensitivity analysis is used to identify the reaction XSs and the dominant chains that contribute most significantly to the uncertainty. The Monte Carlo methodology gives the burnup uncertainty estimates due to the synergetic/global effect of the complete set of XS uncertainties. These uncertainty estimates are valuable to assess the need of any experimental or systematic reevaluation of some uncertainty XSs for ADS.

  17. Sensitivity and Uncertainty Analysis to Burnup Estimates on ADS using the ACAB Code

    SciTech Connect

    Cabellos, O.; Sanz, J.; Rodriguez, A.; Gonzalez, E.; Embid, M.; Alvarez, F.; Reyes, S.

    2005-05-24

    Within the scope of the Accelerator Driven System (ADS) concept for nuclear waste management applications, the burnup uncertainty estimates due to uncertainty in the activation cross sections (XSs) are important regarding both the safety and the efficiency of the waste burning process. We have applied both sensitivity analysis and Monte Carlo methodology to actinides burnup calculations in a lead-bismuth cooled subcritical ADS. The sensitivity analysis is used to identify the reaction XSs and the dominant chains that contribute most significantly to the uncertainty. The Monte Carlo methodology gives the burnup uncertainty estimates due to the synergetic/global effect of the complete set of XS uncertainties. These uncertainty estimates are valuable to assess the need of any experimental or systematic re-evaluation of some uncertainty XSs for ADS.

  18. A study on fast reactor core mechanics by an ex-reactor test and comparisons with calculations

    SciTech Connect

    Tottori, Shoji; Kawanaka, Ikunori; Nakagawa, Masatoshi; Arie, Kazuo; Itoh, Kunihiro; Ohya, Takeaki; Motomiya, Takeo; Adachi, Hironori

    1996-07-01

    This paper presents and discusses the results of core bowing experiments performed with an ex-reactor rig holding a half hexagon array of 22 sub-assemblies (S/As) simulating the Japanese DFBR conditions and the comparisons of the measured results with calculations by individually developed codes--ARKAS, RAINBOW, SANBOW. The main conclusions of this study are (1) interwrapper loads and S/A displacements within the array were measured at selected positions for a series of five tests simulating the DFBR core bowing modes, (2) the overall comparison between the non-friction calculation and measurement showed good agreement for loads, displacements and their directions, and (3) validation of the friction algorithm has also been carried out and further improvement of the agreement was obtained.

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

    NASA Astrophysics Data System (ADS)

    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.

  20. Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics.

    2000-05-12

    Version 00 QUARK is a combined computer program comprising a revised version of the QUANDRY three-dimensional, two-group neutron kinetics code and an upgraded version of the COBRA transient core analysis code (COBRA-EN). Starting from either a critical steady-state (k-effective or critical dilute Boron problem) or a subcritical steady-state (fixed source problem) in a PWR plant, the code allows one to simulate the neutronic and thermal-hydraulic core transient response to reactivity accidents initiated both inside themore » vessel (such as a control rod ejection) and outside the vessel (such as the sudden change of the Boron concentration in the coolant). QUARK output can be used as input to PSR-470/NORMA-FP to perform a subchannel analysis from converged coarse-mesh nodal solutions.« less

  1. Transmission electron microscopy and ab initio calculations to relate interfacial intermixing and the magnetism of core/shell nanoparticles

    SciTech Connect

    Chi, C.-C.; Hsiao, C.-H.; Ouyang, Chuenhou; Skoropata, E.; Lierop, J. van

    2015-05-07

    Significant efforts towards understanding bi-magnetic core-shell nanoparticles are underway currently as they provide a pathway towards properties unavailable with single-phased systems. Recently, we have demonstrated that the magnetism of γ-Fe2O3/CoO core-shell nanoparticles, in particular, at high temperatures, originates essentially from an interfacial doped iron-oxide layer that is formed by the migration of Co{sup 2+} from the CoO shell into the surface layers of the γ-Fe2O3 core [Skoropata et al., Phys. Rev. B 89, 024410 (2014)]. To examine directly the nature of the intermixed layer, we have used high-resolution transmission electron microscopy (HRTEM) and first-principles calculations to examine the impact of the core-shell intermixing at the atomic level. By analyzing the HRTEM images and energy dispersive spectra, the level and nature of intermixing was confirmed, mainly as doping of Co into the octahedral site vacancies of γ-Fe2O3. The average Co doping depths for different processing temperatures (150 °C and 235 °C) were 0.56 nm and 0.78 nm (determined to within 5% through simulation), respectively, establishing that the amount of core-shell intermixing can be altered purposefully with an appropriate change in synthesis conditions. Through first-principles calculations, we find that the intermixing phase of γ-Fe2O3 with Co doping is ferromagnetic, with even higher magnetization as compared to that of pure γ-Fe2O3. In addition, we show that Co doping into different octahedral sites can cause different magnetizations. This was reflected in a change in overall nanoparticle magnetization, where we observed a 25% reduction in magnetization for the 235 °C versus the 150 °C sample, despite a thicker intermixed layer.

  2. Simulating the Dynamics of Earth's Core: Using NCCS Supercomputers Speeds Calculations

    NASA Technical Reports Server (NTRS)

    2002-01-01

    If one wanted to study Earth's core directly, one would have to drill through about 1,800 miles of solid rock to reach liquid core-keeping the tunnel from collapsing under pressures that are more than 1 million atmospheres and then sink an instrument package to the bottom that could operate at 8,000 F with 10,000 tons of force crushing every square inch of its surface. Even then, several of these tunnels would probably be needed to obtain enough data. Faced with difficult or impossible tasks such as these, scientists use other available sources of information - such as seismology, mineralogy, geomagnetism, geodesy, and, above all, physical principles - to derive a model of the core and, study it by running computer simulations. One NASA researcher is doing just that on NCCS computers. Physicist and applied mathematician Weijia Kuang, of the Space Geodesy Branch, and his collaborators at Goddard have what he calls the,"second - ever" working, usable, self-consistent, fully dynamic, three-dimensional geodynamic model (see "The Geodynamic Theory"). Kuang runs his model simulations on the supercomputers at the NCCS. He and Jeremy Bloxham, of Harvard University, developed the original version, written in Fortran 77, in 1996.

  3. No-core Monte Carlo shell model calculations with unitary correlation operator method and similarity renormalization group

    NASA Astrophysics Data System (ADS)

    Liu, Lang

    2015-05-01

    The unitary correlation operator method (UCOM) and the similarity renormalization group theory (SRG) are compared and discussed in the framework of the no-core Monte Carlo shell model (MCSM) calculations for 3H and 4He. The treatment of spurious center-of-mass motion by Lawson's prescription is performed in the MCSM calculations. These results with both transformed interactions show good suppression of spurious center-of-mass motion with proper Lawson's prescription parameter βc.m. values. The UCOM potentials obtain faster convergence of total energy for the ground state than that of SRG potentials in the MCSM calculations, which differs from the cases in the no-core shell model calculations (NCSM). These differences are discussed and analyzed in terms of the truncation scheme in the MCSM and NCSM, as well as the properties of the potentials of SRG and UCOM. Supported by Fundamental Research Funds for the Central Universities (JUSRP1035), National Natural Science Foundation of China (11305077)

  4. Superlinear scaling in master-slave quantum chemical calculations using in-core storage of two-electron integrals.

    PubMed

    Fossgård, Eirik; Ruud, Kenneth

    2006-02-01

    We describe the implementation of a parallel, in-core, integral-direct Hartree-Fock and density functional theory code for the efficient calculation of Hartree-Fock wave functions and density functional theory. The algorithm is based on a parallel master-slave algorithm, and the two-electron integrals calculated by a slave are stored in available local memory. To ensure the greatest computational savings, the master node keeps track of all integral batches stored on the different slaves. The code can reuse undifferentiated two-electron integrals both in the wave function optimization and in the evaluation of second-, third-, and fourth-order molecular properties. Superlinear scaling is achieved in a series of test examples, with speedups of up to 55 achieved for calculations run on medium-sized molecules on 16 processors with respect to the time used on a single processor.

  5. Component mode synthesis methods applied to 3D heterogeneous core calculations, using the mixed dual finite element solver MINOS

    SciTech Connect

    Guerin, P.; Baudron, A. M.; Lautard, J. J.

    2006-07-01

    This paper describes a new technique for determining the pin power in heterogeneous core calculations. It is based on a domain decomposition with overlapping sub-domains and a component mode synthesis technique for the global flux determination. Local basis functions are used to span a discrete space that allows fundamental global mode approximation through a Galerkin technique. Two approaches are given to obtain these local basis functions: in the first one (Component Mode Synthesis method), the first few spatial eigenfunctions are computed on each sub-domain, using periodic boundary conditions. In the second one (Factorized Component Mode Synthesis method), only the fundamental mode is computed, and we use a factorization principle for the flux in order to replace the higher order Eigenmodes. These different local spatial functions are extended to the global domain by defining them as zero outside the sub-domain. These methods are well-fitted for heterogeneous core calculations because the spatial interface modes are taken into account in the domain decomposition. Although these methods could be applied to higher order angular approximations - particularly easily to a SPN approximation - the numerical results we provide are obtained using a diffusion model. We show the methods' accuracy for reactor cores loaded with UOX and MOX assemblies, for which standard reconstruction techniques are known to perform poorly. Furthermore, we show that our methods are highly and easily parallelizable. (authors)

  6. No-Core Shell Model Calculations in Light Nuclei with Three-Nucleon Forces

    SciTech Connect

    Barrett, B R; Vary, J P; Nogga, A; Navratil, P; Ormand, W E

    2004-01-08

    The ab initio No-Core Shell Model (NCSM) has recently been expanded to include nucleon-nucleon (NN) and three-nucleon (3N) interactions at the three-body cluster level. Here it is used to predict binding energies and spectra of p-shell nuclei based on realistic NN and 3N interactions. It is shown that 3N force (3NF) properties can be studied in these nuclear systems. First results show that interactions based on chiral perturbation theory lead to a realistic description of {sup 6}Li.

  7. Three-Dimensional Full Core Power Calculations for Pressurized Water Reactors

    SciTech Connect

    Evans, Thomas M; Davidson, Gregory G; Slaybaugh, Rachel N

    2010-01-01

    We have implemented a new multilevel parallel decomposition in the Denovo discrete ordinates radiation transport code. In concert with Krylov subspace iterative solvers, the multilevel decomposition allows concurrency over energy in addition to space-angle. The original space-angle partitioning in Denovo placed an eective limit on the scalability of the transport solver that was highly dependent on the problem size. The added phase-space concurrency combined with the high-performance Krylov solvers has enabled weak scaling to 100K cores on the Jaguar XT5 supercomputer. Furthermore, the multilevel decomposition provides enough concurrency to scale to exascale computing and beyond.

  8. Weak-coupling calculations in the /sup 208/Pb core region

    SciTech Connect

    McGrory, J.B.

    1981-01-01

    The structure of Tl and Hg isotopes near the /sup 208/Pb core is discussed in terms of a weak-coupling model where low-lying proton-hole states are coupled to low-lying neutron-hole states. The even Pb isotopes are first discussed in terms of a generalized seniority-2 approximation which is shown to be very accurate. The weak-coupling results are not satisfactory, and it is suggested that the defeat may be in the n-p interaction.

  9. Revised Burnup Code System SWAT: Description and Validation Using Postirradiation Examination Data

    SciTech Connect

    Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

    2002-05-15

    The burnup code system Step-Wise Burnup Analysis Code System (SWAT) is revised for use in a burnup credit analysis. An important feature of the revised SWAT is that its functions are achieved by calling validated neutronics codes without any changes to the original codes. This feature is realized with a system function of the operating system, which allows the revised SWAT to be independent of the development status of each code.A package of the revised SWAT contains the latest libraries based on JENDL-3.2 and the second version of the JNDC FP library. These libraries allow us to analyze burnup problems, such as an analysis of postirradiation examination (PIE), using the latest evaluated data of not only cross sections but also fission yield and decay constants.Another function of the revised SWAT is a library generator for the ORIGEN2 code, which is one of the most reliable burnup codes. ORIGEN2 users can obtain almost the same results with the revised SWAT using the library prepared by this function.The validation of the revised SWAT is conducted by calculation of the Organization for Economic Cooperation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark Phase I-B and analyses of PIE data for spent fuel from Takahama Unit 3. The analysis of PIE data shows that the revised SWAT can predict the isotopic composition of main uranium and plutonium with a deviation of 5% from experimental results taken from UO{sub 2} fuels of 17 x 17 fuel assemblies. Many results of fission products including samarium are within a deviation of 10%. This means that the revised SWAT has high reliability to predict the isotopic composition for pressurized water reactor spent fuel.

  10. An Automated, Multi-Step Monte Carlo Burnup Code System.

    2003-07-14

    Version 02 MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2,more » the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included.« less

  11. An Automated, Multi-Step Monte Carlo Burnup Code System.

    SciTech Connect

    TRELLUE, HOLLY R.

    2003-07-14

    Version 02 MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2, the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included.

  12. Reactor physics calculations for {sup 99}Mo production at the annular core research reactor

    SciTech Connect

    Parma, E.J.

    1995-12-31

    The Isotope Production and Distribution Program at the U.S. Department of Energy has designated Sandia National Laboratories (SNL) as the most appropriate facility for the production of {sup 99}Mo, a radioisotope whose daughter, {sup 99m}Tc, is used in more than 36,000 medical procedures per day in the United States and is considered to be a vital medical diagnostic and treatment tool. The isotope would be produced at SNL using the annular core research reactor (ACRR) facility and collocated hot cell facility. The {sup 99}Mo would be produced using the fission process by irradiating {open_quotes}targets{close_quotes} coated with {sup 235}U in the form of highly enriched U{sub 3}O{sub 8}. After {approximately}7 days of continuous irradiation in the ACRR, a target would be re- moved from the reactor core for processing. The isotope would be extracted by chemically precipitating the molybdenum using the {open_quotes}Cintichem{close_quotes} process and would be shipped to the various pharmaceutical companies by commercial or chartered airline.

  13. AB initio free energy calculations of the solubility of silica in metallic hydrogen and application to giant planet cores

    SciTech Connect

    González-Cataldo, F.; Wilson, Hugh F.; Militzer, B.

    2014-05-20

    By combining density functional molecular dynamics simulations with a thermodynamic integration technique, we determine the free energy of metallic hydrogen and silica, SiO{sub 2}, at megabar pressures and thousands of degrees Kelvin. Our ab initio solubility calculations show that silica dissolves into fluid hydrogen above 5000 K for pressures from 10 and 40 Mbars, which has implications for the evolution of rocky cores in giant gas planets like Jupiter, Saturn, and a substantial fraction of known extrasolar planets. Our findings underline the necessity of considering the erosion and redistribution of core materials in giant planet evolution models, but they also demonstrate that hot metallic hydrogen is a good solvent at megabar pressures, which has implications for high-pressure experiments.

  14. Intrinsic electrostatic resonances of heterostructures with negative permittivity from finite-element calculations: Application to core-shell inclusions

    NASA Astrophysics Data System (ADS)

    Mejdoubi, Abdelilah; Brosseau, Christian

    2007-11-01

    Herein, we report finite-element calculations of the effective (relative) permittivity of composite materials consisting of inclusions and inclusion arrays with a core-shell structure embedded in a surrounding host. The material making up the core of the two-dimensional structures, or cross sections of infinite three-dimensional objects (parallel, infinitely long, and identical cylinders) where the properties and characteristics are invariant along the perpendicular cross sectional plane, is assumed to have a negative real part of the permittivity, while the coating material (annular shell) is considered to be lossless. While strictly valid only in a dc situation, our analysis can be extended to treat electric fields that oscillate with time, provided that the wavelengths and attenuation lengths associated with the fields are much larger than the microstructure dimension in order that the homogeneous (effective-medium) representation of the composite structure makes sense. While one may identify features of the electrostatic resonance (ER) which are common to core-shell structures characterized by permittivities with real parts of opposite signs, it appears that the predicted ER positions are sensitive to the shell thickness and can be tuned through varying this geometric parameter. For example, we observe that the ER is broadened and shifted as the loss and the shell thickness are increased, respectively. We also argue that such core shell may also be valuable in controlling ER characteristics via polarization in an external electric field. In addition, by considering calculations of the electric field distribution, we find that the ER results in very strong and local-field enhancements into small parts of the shell perimeter. Our findings open up possibilities for the development of hybrid structures that could exploit the ER features for a particular application.

  15. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

    SciTech Connect

    Wagner, J.C.

    2001-09-28

    The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs

  16. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    NASA Astrophysics Data System (ADS)

    Makmal, T.; Aviv, O.; Gilad, E.

    2016-10-01

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.

  17. Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    SciTech Connect

    Ade, Brian J; Bowman, Stephen M; Gauld, Ian C; Ilas, Germina; Martinez, J. S.

    2015-01-01

    [Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, and it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades

  18. Three Dimensional Analysis of 3-Loop PWR RCCA Ejection Accident for High Burnup

    SciTech Connect

    Marciulescu, Cristian; Sung, Yixing; Beard, Charles L.

    2006-07-01

    The Rod Control Cluster Assembly (RCCA) ejection accident is a Condition IV design basis reactivity insertion event for Pressurized Water Reactors (PWR). The event is historically analyzed using a one-dimensional (1D) neutron kinetic code to meet the current licensing criteria for fuel rod burnup to 62,000 MWD/MTU. The Westinghouse USNRC-approved three-dimensional (3D) analysis methodology is based on the neutron kinetics version of the ANC code (SPNOVA) coupled with Westinghouse's version of the EPRI core thermal-hydraulic code VIPRE-01. The 3D methodology provides a more realistic yet conservative analysis approach to meet anticipated reduction in the licensing fuel enthalpy rise limit for high burnup fuel. A rod ejection analysis using the 3D methodology was recently performed for a Westinghouse 3-loop PWR at an up-rated core power of 3151 MWt with reload cores that allow large flexibility in assembly shuffling and a fuel hot rod burnup to 75,000 MWD/MTU. The analysis considered high enrichment fuel assemblies at the control rod locations as well as bounding rodded depletions in the end of life, zero power and full power conditions. The analysis results demonstrated that the peak fuel enthalpy rise is less than 100 cal/g for the transient initiated at the hot zero power condition. The maximum fuel enthalpy is less than 200 cal/g for the transient initiated from the full power condition. (authors)

  19. Parametric Study of Control Rod Exposure for PWR Burnup Credit Criticality Safety Analyses

    SciTech Connect

    Sanders, C.E.

    2001-07-20

    The Interim Staff Guidance on burnup credit (ISG-8) for pressurized water reactor (PWR) spent nuclear fuel (SNF), issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office, recommends the use of analyses that provide an ''adequate representation of the physics'' and notes particular concern with the ''need to consider the more reactive actinide compositions of fuels burned with fixed absorbers or with control rods fully or partly inserted.'' In the absence of readily available information on the extent of control rod (CR) usage in U.S. PWRs and the subsequent reactivity effect of CR exposure on discharged SNF, NRC staff have indicated a need for greater understanding in these areas. In response, this paper presents results of a parametric study of the effect of CR exposure on the reactivity of discharged SNF for various CR designs (including Axial Power Shaping Rods), fuel enrichments, and exposure conditions (i.e., burnup and axial insertion). The study is performed in two parts. In the first part, two-dimensional calculations are performed, effectively assuming full axial CR insertion. These calculations are intended to bound the effect of CR exposure and facilitate comparisons of the various CR designs. In the second part, three-dimensional calculations are performed to determine the effect of various axial insertion conditions and gain a better understanding of reality. The results from the study demonstrate that the reactivity effect increases with increasing CR exposure (e.g., burnup) and decreasing initial fuel enrichment (for a fixed burnup). Additionally, the results show that even for significant burnup exposures, minor axial CR insertions (e.g., < 20 cm) result in an insignificant effect on the k{sub eff} of a spent fuel cask.

  20. High burnup effects in WWER fuel rods

    SciTech Connect

    Smirnov, V.; Smirnov, A.

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

  1. Density functional theory calculation on many-cores hybrid central processing unit-graphic processing unit architectures.

    PubMed

    Genovese, Luigi; Ospici, Matthieu; Deutsch, Thierry; Méhaut, Jean-François; Neelov, Alexey; Goedecker, Stefan

    2009-07-21

    We present the implementation of a full electronic structure calculation code on a hybrid parallel architecture with graphic processing units (GPUs). This implementation is performed on a free software code based on Daubechies wavelets. Such code shows very good performances, systematic convergence properties, and an excellent efficiency on parallel computers. Our GPU-based acceleration fully preserves all these properties. In particular, the code is able to run on many cores which may or may not have a GPU associated, and thus on parallel and massive parallel hybrid machines. With double precision calculations, we may achieve considerable speedup, between a factor of 20 for some operations and a factor of 6 for the whole density functional theory code.

  2. Ab initio no core calculations of light nuclei and preludes to Hamiltonian quantum field theory

    SciTech Connect

    Vary, J.P.; Maris, P.; Shirokov, A.M.; Honkanen, H.; li, J.; Brodsky, S.J.; Harindranath, A.; Teramond, G.F.de; /Costa Rica U.

    2009-08-03

    Recent advances in ab initio quantum many-body methods and growth in computer power now enable highly precise calculations of nuclear structure. The precision has attained a level sufficient to make clear statements on the nature of 3-body forces in nuclear physics. Total binding energies, spin-dependent structure effects, and electroweak properties of light nuclei play major roles in pinpointing properties of the underlying strong interaction. Eventually,we anticipate a theory bridge with immense predictive power from QCD through nuclear forces to nuclear structure and nuclear reactions. Light front Hamiltonian quantum field theory offers an attractive pathway and we outline key elements.

  3. Insights Into the PA Neutral Atom: from AN Evaluation of PA2+ Outer-Core Correlation Energy Calculations

    NASA Astrophysics Data System (ADS)

    Mrozik, Michael K.; Pitzer, Russell M.; Bursten, Bruce E.

    2010-06-01

    Since the identification of f-orbital contribution to the bonding in PaO+, investigations into Pa cations have hoped to characterize as many of the electronic states possible.1 Electronic states of the Pan+ (n=0-4) ions have been investigated using multi-reference spin-orbit configuration interaction (MR-SOCI). Initial investigations using Dunning style correlation consistent double-{ζ} basis sets are re-examined with a larger triple-{ζ} basis, with the hope of supporting the order of electronic states. Calculations using Hartree-Fock and CI calculations on the neutral atom did not produce the known order of states. A case study was deemed necessary on similar electron configurations present in the low energy states of Pa2+ more specifically those generated from the 5f26d1 and 5f16d2 configurations. Comparison in the Pa2+ ion is complicated by the lack of experimental results, but the states are presumed to be similar sequence as those in the neutral atom, with the addition of two electrons in the 7s shell. In evaluating the impact of inclusion of the outer core, calculations including valence-outer core correlation were completed for the 5d, 6s, and 6p shells of the Pa2+ ion. The magnitude of these individual shell correlation calculations will allow for identification of the energy level shifts associated with even and odd configurations, better describing the energy order in both the Pa2+ ion case study and for the neutral Pa atom. Upon completion of this aspect of the Pa neutral atom study, the knowledge of the energy levels in the Pan+ (n=0-4) family of ions will be greatly expanded, and may yield a model for future studies of atomic actinide systems. Gibson {et al.} Organometallics 2007, 26, 3947-3956.

  4. Short period ScP phase amplitude calculations for core-mantle boundary with intermediate scale topography

    NASA Astrophysics Data System (ADS)

    Shen, Zhichao; Ni, Sidao; Wu, Wenbo; Sun, Daoyuan

    2016-04-01

    The core-mantle boundary (CMB) topography plays a key role in constraining geodynamic modeling and core-mantle coupling. It's effective to resolve the intermediate lateral scale topography (hundreds of km) with short period core reflected seismic phases (ScP) due to their small Fresnel-zones at short epicentral distances (<3336 km (30°)). We developed a method based on the ray theory and representation theorem to calculate short period ScP synthetics for intermediate lateral scale CMB topography. The CMB topography we introduced here is axisymmetric and specified with two parameters: H (height) and L (diameter, or lateral length scale). Our numerical computation shows that a bump (H > 0) and dip (H < 0) model would cause defocusing/weakening and focusing/amplifying effects on ScP amplitude. Moreover, the effect of frequency and combination of L and H are quantified with the amplification coefficients. Then we applied this method to estimate a possible CMB topography beneath northeastern Japan, and a CMB model with L = 140 km, H = 1.2 km overall matches the observed pattern of 2D PcP/ScP amplitude ratios. However, it is difficult to totally rule out other factors that may also affect PcP/ScP pattern because of limitation of ray-based algorithms we used here. A hybrid method combining ray theory and numerical method is promising for studying complicated 3D structure and CMB topography in the future.

  5. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    SciTech Connect

    Marshall, William BJ J

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  6. First-principles calculations of properties of orthorhombic iron carbide Fe7C3 at the Earth's core conditions

    NASA Astrophysics Data System (ADS)

    Raza, Zamaan; Shulumba, Nina; Caffrey, Nuala M.; Dubrovinsky, Leonid; Abrikosov, Igor A.

    2015-06-01

    A recently discovered phase of orthorhombic iron carbide o-Fe7C3 [Prescher et al., Nat. Geosci. 8, 220 (2015), 10.1038/ngeo2370] is assessed as a potentially important phase for interpretation of the properties of the Earth's core. In this paper, we carry out first-principles calculations on o-Fe7C3 , finding properties to be in broad agreement with recent experiments, including a high Poisson's ratio (0.38). Our enthalpy calculations suggest that o-Fe7C3 is more stable than Eckstrom-Adcock hexagonal iron carbide (h-Fe7C3 ) below approximately 100 GPa. However, at 150 GPa, the two phases are essentially degenerate in terms of Gibbs free energy, and further increasing the pressure towards Earth's core conditions stabilizes h-Fe7C3 with respect to the orthorhombic phase. Increasing the temperature tends to stabilize the hexagonal phase at 360 GPa, but this trend may change beyond the limit of the quasiharmonic approximation.

  7. Development and testing of a compact basis set for use in effective core potential calculations on rhodium complexes.

    PubMed

    Roscioni, Otello M; Lee, Edmond P F; Dyke, John M

    2012-10-01

    We present a set of effective core potential (ECP) basis sets for rhodium atoms which are of reasonable size for use in electronic structure calculations. In these ECP basis sets, the Los Alamos ECP is used to simulate the effect of the core electrons while an optimized set of Gaussian functions, which includes polarization and diffuse functions, is used to describe the valence electrons. These basis sets were optimized to reproduce the ionization energy and electron affinity of atomic rhodium. They were also tested by computing the electronic ground state geometry and harmonic frequencies of [Rh(CO)(2) μ-Cl](2) , Rh(CO)(2) ClPy, and RhCO (neutral and its positive, and negative ions) as well as the enthalpy of the reaction of [Rh(CO)(2) μ-Cl](2) with pyridine (Py) to give Rh(CO)(2) ClPy, at different levels of theory. Good agreement with experimental values was obtained. Although the number of basis functions used in our ECP basis sets is smaller than those of other ECP basis sets of comparable quality, we show that the newly developed ECP basis sets provide the flexibility and precision required to reproduce a wide range of chemical and physical properties of rhodium compounds. Therefore, we recommend the use of these compact yet accurate ECP basis sets for electronic structure calculations on molecules involving rhodium atoms.

  8. Benefits of the delta K of depletion benchmarks for burnup credit validation

    SciTech Connect

    Lancaster, D.; Machiels, A.

    2012-07-01

    Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO{sub 2} critical experiments to determine the criticality safety limits on the neutron multiplication factor, k{sub eff}. The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

  9. Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

    SciTech Connect

    Wagner, J.C.

    2002-12-17

    This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

  10. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    SciTech Connect

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  11. Chemical solver to compute molecule and grain abundances and non-ideal MHD resistivities in prestellar core-collapse calculations

    NASA Astrophysics Data System (ADS)

    Marchand, P.; Masson, J.; Chabrier, G.; Hennebelle, P.; Commerçon, B.; Vaytet, N.

    2016-07-01

    We develop a detailed chemical network relevant to calculate the conditions that are characteristic of prestellar core collapse. We solve the system of time-dependent differential equations to calculate the equilibrium abundances of molecules and dust grains, with a size distribution given by size-bins for these latter. These abundances are used to compute the different non-ideal magneto-hydrodynamics resistivities (ambipolar, Ohmic and Hall), needed to carry out simulations of protostellar collapse. For the first time in this context, we take into account the evaporation of the grains, the thermal ionisation of potassium, sodium, and hydrogen at high temperature, and the thermionic emission of grains in the chemical network, and we explore the impact of various cosmic ray ionisation rates. All these processes significantly affect the non-ideal magneto-hydrodynamics resistivities, which will modify the dynamics of the collapse. Ambipolar diffusion and Hall effect dominate at low densities, up to nH = 1012 cm-3, after which Ohmic diffusion takes over. We find that the time-scale needed to reach chemical equilibrium is always shorter than the typical dynamical (free fall) one. This allows us to build a large, multi-dimensional multi-species equilibrium abundance table over a large temperature, density and ionisation rate ranges. This table, which we make accessible to the community, is used during first and second prestellar core collapse calculations to compute the non-ideal magneto-hydrodynamics resistivities, yielding a consistent dynamical-chemical description of this process. The multi-dimensional multi-species equilibrium abundance table and a copy of the code are only available at the CDS via anonymous ftp to http://cdsarc.u-strasbg.fr (http://130.79.128.5) or via http://cdsarc.u-strasbg.fr/viz-bin/qcat?J/A+A/592/A18

  12. Analysis and comparison of CVS-ADC approaches up to third order for the calculation of core-excited states

    SciTech Connect

    Wenzel, Jan Holzer, Andre; Wormit, Michael; Dreuw, Andreas

    2015-06-07

    The extended second order algebraic-diagrammatic construction (ADC(2)-x) scheme for the polarization operator in combination with core-valence separation (CVS) approximation is well known to be a powerful quantum chemical method for the calculation of core-excited states and the description of X-ray absorption spectra. For the first time, the implementation and results of the third order approach CVS-ADC(3) are reported. Therefore, the CVS approximation has been applied to the ADC(3) working equations and the resulting terms have been implemented efficiently in the adcman program. By treating the α and β spins separately from each other, the unrestricted variant CVS-UADC(3) for the treatment of open-shell systems has been implemented as well. The performance and accuracy of the CVS-ADC(3) method are demonstrated with respect to a set of small and middle-sized organic molecules. Therefore, the results obtained at the CVS-ADC(3) level are compared with CVS-ADC(2)-x values as well as experimental data by calculating complete basis set limits. The influence of basis sets is further investigated by employing a large set of different basis sets. Besides the accuracy of core-excitation energies and oscillator strengths, the importance of cartesian basis functions and the treatment of orbital relaxation effects are analyzed in this work as well as computational timings. It turns out that at the CVS-ADC(3) level, the results are not further improved compared to CVS-ADC(2)-x and experimental data, because the fortuitous error compensation inherent in the CVS-ADC(2)-x approach is broken. While CVS-ADC(3) overestimates the core excitation energies on average by 0.61% ± 0.31%, CVS-ADC(2)-x provides an averaged underestimation of −0.22% ± 0.12%. Eventually, the best agreement with experiments can be achieved using the CVS-ADC(2)-x method in combination with a diffuse cartesian basis set at least at the triple-ζ level.

  13. Evaluation of accuracy of calculations of VVER-1000 core states with incomplete covering of fuel by the absorber

    SciTech Connect

    Tikhomirov, A. V.; Ponomarenko, G. L.

    2012-07-01

    An additional verification of bundled software (BS) SAPFIR-95 and amp;RC [1] and code KORSAR/GP [2] was performed. Both software products were developed in A.P. Alexandrov NITI and certified by ROSTEKHNADZOR of RF for numeric simulation of stationary, transitional and emergency conditions of VVER reactors. A benchmark model for neutronics calculations was created within the limits of this work. The cold subcritical state of VVER - 1000 reactor stationary fuelling was simulated on the basis of FA with an increased height of the fuel column (TVS-2M) considering detailed presentation of radial and front neutron reflectors. A case of passing of pure condensate slug through the core in initially deep subcritical state during start of the first RCP set after refueling was considered as an examined condition of reactor operation. A relatively small size of the slug, its spatial position near the reflectors (lower and lateral), as well as failure of the inserted control rods of the control and protection system (CPS CR) to reach the lower limit of the fuel column stipulate for methodical complexity of a correct calculation of the neutron multiplication constant (K{sub eff}) using engineering codes. Code RC was used as a test program in the process of reactor calculated 3-D modeling. Code MCNP5 [3] was used as the precision program, which solves the equation of neutrons transfer by Monte-Carlo method and which was developed in the US (Los-Alamos). As a result of comparative calculations dependency of K{sub eff} on two parameters was evaluated - boron acid concentration (Cb) and CPS CR position. Reactivity effect was evaluated, which is implemented as a result of failure of all CPS control rods to reach the lower fuel limit calculated using the engineering codes mentioned above. (authors)

  14. Performance of heterogeneous computing with graphics processing unit and many integrated core for hartree potential calculations on a numerical grid.

    PubMed

    Choi, Sunghwan; Kwon, Oh-Kyoung; Kim, Jaewook; Kim, Woo Youn

    2016-09-15

    We investigated the performance of heterogeneous computing with graphics processing units (GPUs) and many integrated core (MIC) with 20 CPU cores (20×CPU). As a practical example toward large scale electronic structure calculations using grid-based methods, we evaluated the Hartree potentials of silver nanoparticles with various sizes (3.1, 3.7, 4.9, 6.1, and 6.9 nm) via a direct integral method supported by the sinc basis set. The so-called work stealing scheduler was used for efficient heterogeneous computing via the balanced dynamic distribution of workloads between all processors on a given architecture without any prior information on their individual performances. 20×CPU + 1GPU was up to ∼1.5 and ∼3.1 times faster than 1GPU and 20×CPU, respectively. 20×CPU + 2GPU was ∼4.3 times faster than 20×CPU. The performance enhancement by CPU + MIC was considerably lower than expected because of the large initialization overhead of MIC, although its theoretical performance is similar with that of CPU + GPU. © 2016 Wiley Periodicals, Inc. PMID:27431905

  15. Performance of heterogeneous computing with graphics processing unit and many integrated core for hartree potential calculations on a numerical grid.

    PubMed

    Choi, Sunghwan; Kwon, Oh-Kyoung; Kim, Jaewook; Kim, Woo Youn

    2016-09-15

    We investigated the performance of heterogeneous computing with graphics processing units (GPUs) and many integrated core (MIC) with 20 CPU cores (20×CPU). As a practical example toward large scale electronic structure calculations using grid-based methods, we evaluated the Hartree potentials of silver nanoparticles with various sizes (3.1, 3.7, 4.9, 6.1, and 6.9 nm) via a direct integral method supported by the sinc basis set. The so-called work stealing scheduler was used for efficient heterogeneous computing via the balanced dynamic distribution of workloads between all processors on a given architecture without any prior information on their individual performances. 20×CPU + 1GPU was up to ∼1.5 and ∼3.1 times faster than 1GPU and 20×CPU, respectively. 20×CPU + 2GPU was ∼4.3 times faster than 20×CPU. The performance enhancement by CPU + MIC was considerably lower than expected because of the large initialization overhead of MIC, although its theoretical performance is similar with that of CPU + GPU. © 2016 Wiley Periodicals, Inc.

  16. Hybrid density functional-molecular mechanics calculations for core-electron binding energies of glycine in water solution.

    PubMed

    Niskanen, Johannes; Arul Murugan, N; Rinkevicius, Zilvinas; Vahtras, Olav; Li, Cui; Monti, Susanna; Carravetta, Vincenzo; Agren, Hans

    2013-01-01

    We report hybrid density functional theory-molecular mechanics (DFT/MM) calculations performed for glycine in water solution at different pH values. In this paper, we discuss several aspects of the quantum mechanics-molecular mechanics (QM/MM) simulations where the dynamics and spectral binding energy shifts are computed sequentially, and where the latter are evaluated over a set of configurations generated by molecular or Car-Parrinello dynamics simulations. In the used model, core ionization takes place in glycine as a quantum mechanical (QM) system modeled with DFT, and the solution is described with expedient force fields in a large molecular mechanical (MM) volume of water molecules. The contribution to the core electronic binding energy from all interactions within and between the two (DFT and MM) parts is accounted for, except charge transfer and dispersion. While the obtained results were found to be in qualitative agreement with experiment, their precision must be qualified with respect to the problem of counter ions, charge transfer and optimal division of QM and MM parts of the system. Results are compared to those of a recent study [Ottoson et al., J. Am. Chem. Soc., 2011, 133, 3120].

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

    SciTech Connect

    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

  18. Microstructural Modeling of Thermal Conductivity of High Burn-up Mixed Oxide Fuel

    SciTech Connect

    Melissa Teague; Michael Tonks; Stephen Novascone; Steven Hayes

    2014-01-01

    Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez–Lucuta model was favorable.

  19. Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel

    NASA Astrophysics Data System (ADS)

    Teague, Melissa; Tonks, Michael; Novascone, Stephen; Hayes, Steven

    2014-01-01

    Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON [1] fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez-Lucuta model was favorable.

  20. Determination of deuterium–tritium critical burn-up parameter by four temperature theory

    SciTech Connect

    Nazirzadeh, M.; Ghasemizad, A.; Khanbabei, B.

    2015-12-15

    Conditions for thermonuclear burn-up of an equimolar mixture of deuterium-tritium in non-equilibrium plasma have been investigated by four temperature theory. The photon distribution shape significantly affects the nature of thermonuclear burn. In three temperature model, the photon distribution is Planckian but in four temperature theory the photon distribution has a pure Planck form below a certain cut-off energy and then for photon energy above this cut-off energy makes a transition to Bose-Einstein distribution with a finite chemical potential. The objective was to develop four temperature theory in a plasma to calculate the critical burn up parameter which depends upon initial density, the plasma components initial temperatures, and hot spot size. All the obtained results from four temperature theory model are compared with 3 temperature model. It is shown that the values of critical burn-up parameter calculated by four temperature theory are smaller than those of three temperature model.

  1. S∧4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients

    NASA Astrophysics Data System (ADS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The S∧4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at ~1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S∧4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 ¢/K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S∧4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 ¢/K -0.9073 and 0.8502 $/atom%) are the lowest.

  2. Evaluation of Fission Product Critical Experiments and Associated Biases for Burnup Credit Validation

    SciTech Connect

    Mueller, Don; Rearden, Bradley T; Reed, Davis Allan

    2010-01-01

    One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets of critical experiments that are similar to the criticality safety models. A principal challenge for crediting fission products (FP) in a burnup credit safety evaluation is the limited availability of relevant FP critical experiments for bias and bias uncertainty determination. This paper provides an evaluation of the available critical experiments that include FPs, along with bounding, burnup-dependent estimates of FP biases generated by combining energy dependent sensitivity data for a typical burnup credit application with the nuclear data uncertainty information distributed with SCALE 6. A method for determining separate bias and bias uncertainty values for individual FPs and illustrative results is presented. Finally, a FP bias calculation method based on data adjustment techniques and reactivity sensitivity coefficients calculated with the SCALE sensitivity/uncertainty tools and some typical results is presented. Using the methods described in this paper, the cross-section bias for a representative high-capacity spent fuel cask associated with the ENDF/B-VII nuclear data for 16 most important stable or near stable FPs is predicted to be no greater than 2% of the total worth of the 16 FPs, or less than 0.13 % k/k.

  3. Using NDA Techniques to Improve Safeguards Metrics on Burnup Quantification and Plutonium Content in LWR SNF

    SciTech Connect

    Saavedra, Steven F; Charlton, William S; Solodov, Alexander A; Ehinger, Michael H

    2010-01-01

    Globally, there exists a long history in reprocessing in evaluation of the shipper/receiver difference (SRD) on spent nuclear fuel (SNF) received and processed. Typically, the declared shipper s values for uranium and plutonium in SNF (based on calculations involving the initial manufacturer s data and reactor operating history) are used as the input quantities to the head-end process of the facility. Problems have been encountered when comparing these values with measured results of the input accountability tank contents. A typical comparison yields a systematic bias indicated as a loss of 5 7 percent of the plutonium (Pu) and approximately 1 percent for the uranium (U). Studies suggest that such deviation can be attributed to the non-linear nature of the axial burnup values of the SNF. Oak Ridge National Laboratory and Texas A&M University are co-investigating the development of a new method, via Nondestructive Assay (NDA) techniques, to improve the accuracy in burnup and Pu content quantification. Two major components have been identified to achieve this objective. The first component calculates a measurement-based burnup profile along the axis of a fuel rod. Gamma-ray data is collected at numerous locations along the axis of the fuel rod using a High Purity Germanium (HPGe) detector designed for a wide range of gamma-ray energies. Using two fission products, 137Cs and 134Cs, the burnup is calculated at each measurement location and a profile created along the axis of the rod based on the individual measurement locations. The second component measures the U/Pu ratio using an HPGe detector configured for relatively low-energy gamma-rays including x-rays. Fluorescence x-rays from U and Pu are measured and compared to the U/Pu ratio determined from a destructive analysis of the sample. This will be used to establish a relationship between the measured and actual values. This relationship will be combined with the burnup analysis results to establish a relationship

  4. Analysis of burnup and isotopic compositions of BWR 9 x 9 UO{sub 2} fuel assemblies

    SciTech Connect

    Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.

    2012-07-01

    In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO{sub 2} fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for {sup 238}Pu, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148}Nd, {sup 134}Cs, {sup 154}Eu, {sup 152}Sm, {sup 154}Gd, and {sup 157}Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

  5. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

    PubMed

    Lashkari, A; Khalafi, H; Kazeminejad, H

    2013-05-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

  6. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

    PubMed

    Lashkari, A; Khalafi, H; Kazeminejad, H

    2013-05-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

  7. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    PubMed Central

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

  8. Uncertainties in the effects of burnup and their impact on criticality safety licensing criteria

    SciTech Connect

    Carlson, R.W.; Fisher, L.E.

    1990-07-13

    Current criteria for criticality safety for spent fuel shipping and storage casks are conservative because no credit is permitted for the effects of burnup of the fuel inside the cask. Cask designs that will transport and store large numbers of fuel assemblies (20 or more) must devote a substantial part of their payload to criticality control measures if they are to meet this criteria. The Department of Energy is developing the data necessary to support safety analyses that incorporate the effects of burnup for the next generation of spent fuel shipping casks. The efforts described here are devoted to the development of acceptance criteria that will be the basis for accepting safety analyses. Preliminary estimates of the uncertainties of the effects of burnup have been developed to provide a basis for the consideration of critically safety criteria. The criticality safety margins in a spent fuel shipping or storage cask are dominated by the portions of a fuel assembly that are in low power regions of a reactor core, and the reactor operating conditions are very different from spent fuel storage or transport cask conditions. Consequently, the experience that has been gathered during years of reactor operation does not apply directly to the prediction of criticality safety margins for spent fuel shipping or storage casks. The preliminary estimates of the uncertainties presented in this paper must be refined by both analytical and empirical studies that address both the magnitude of the uncertainties and their interdependence. 9 refs., 5 figs.

  9. Preparation of higher-actinide burnup and cross section samples. [LMFBR

    SciTech Connect

    Adair, H.L.; Kobisk, E.H.; Quinby, T.C.; Thomas, D.K.; Dailey, J.M.

    1981-01-01

    A joint research program involving the United States and the United Kingdom was instigated about four years ago for the purpose of studying burnup of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of cross sections of a wide variety of higher actinide isotopes was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the burnup and cross section samples. The higher actinide samples chosen for the burnup study were /sup 241/Am and /sup 244/Cm in the forms of Am/sub 2/O/sub 3/, Cm/sub 2/O/sub 3/, and Am/sub 6/ Cm(RE)/sub 7/O/sub 21/, where (RE) represents a mixture of lanthanide sesquioxides. It is the purpose of this paper to describe technology development and its application in the preparation of the fuel specimens and the cross section specimens that are being used in this cooperative program.

  10. DANDE: a linked code system for core neutronics/depletion analysis

    SciTech Connect

    LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

    1985-06-01

    This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem.

  11. Unexpectedly strong energy stabilization inside the hydrophobic core of small protein rubredoxin mediated by aromatic residues: correlated ab initio quantum chemical calculations.

    PubMed

    Vondrásek, Jirí; Bendová, Lada; Klusák, Vojtech; Hobza, Pavel

    2005-03-01

    The formation of a hydrophobic core of globular proteins is believed to be the consequence of exterior hydrophobic forces of entropic nature. This, together with the low occurrence of hydrogen bonds in the protein core, leads to the opinion that the energy contribution of core formation to protein folding and stability is negligible. We show that stabilization inside the hydrophobic core of a small protein, rubredoxin, determined by means of high-level correlated ab initio calculations (complete basis set limit of MP2 stabilization energy + CCSD(T) correction term), amounted to approximately 50 kcal/mol. These results clearly demonstrate strong attraction inside a hydrophobic core. This finding may lead to substantial changes in the current view of protein folding. We also point out the inability of the DFT/B3LYP method to describe a strong attraction between studied amino acids.

  12. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    SciTech Connect

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  13. Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

    SciTech Connect

    Wagner, J.C.

    2001-08-02

    This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.

  14. Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

    SciTech Connect

    G. S. Chang

    2005-08-01

    A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data.

  15. A feasibility study to determine cooling time and burnup of ATR fuel using a nondestructive technique and three types of gamma-ray detectors

    SciTech Connect

    Navarro, J.; Aryaeinejad, R.; Nigg, D.W.

    2011-07-01

    The goal of this work was to perform a feasibility study and establish measurement techniques to determine the burnup of the Advanced Test Reactor (ATR) fuels at the Idaho National Laboratory (INL). Three different detectors of high purity germanium (HPGe), lanthanum bromide (LaBr{sub 3}), and high pressure xenon (HPXe) in two detection system configurations of below and above the water pool were used in this study. The last two detectors were used for the first time in fuel burnup measurements. The results showed that a better quality spectra can be achieved with the above the water pool configuration. Both short and long cooling time fuels were investigated in order to determine which measurement technique, absolute or fission product ratio, is better suited in each scenario and also to establish what type of detector should be used in each case for the best burnup measurement. The burnup and cooling time calibrations were established using experimental absolute activities or isotopic ratios and ORIGEN burnup calculations. A method was developed to do burnup and cooling time calibrations using fission isotopes activities without the need to know the exact geometry. (authors)

  16. Evaluation of burnup credit for fuel storage analysis -- Experience in Spain

    SciTech Connect

    Conde, J.M.; Recio, M.

    1995-04-01

    Several Spanish light water reactor commercial nuclear power plants are close to maximum spent-fuel pool storage capacity. The utilities are working on the implementation of state-of-the-art methods to increase the storage capacity, including both changes in the pool design (recracking) and the implementation of new analysis approaches with reduced conservation (burnup credit). Burnup credit criticality safety analyses have been approved for two pressurized water reactor plants (four units) and one boiling water reactor (BWR); an other BWR storage analysis is being developed at this moment. The elimination of the ``fresh fuel assumption`` increases the complexity of the criticality analysis to be performed, sometimes putting into question the capability of the analytic tools to properly describe this new situation and increasing the scope of the scenarios to be analyzed. From a regulatory perspective, the reactivity reduction associated with burnup of the fuel can be given credit only if the exposure of each fuel bundle can be known with enough accuracy. Subcriticality of spent-fuel storage depends mainly on the initial fuel enrichment, storage geometry, fuel exposure history, and cooling time. The last two aspects introduced new uncertainties in the criticality analysis that should be quantified in an adequate way. In addition, each and every fuel bundle has its own specific exposure history, so that strong assumptions and simplified calculational schemes have to be developed to undertake the analysis. The Consejo de Seguridad Nuclear (CSN), Spanish regulatory authority on the matter of nuclear safety and radiation protection, plays an active role in the development of analysis methods to support burnup credit, making proposals that may be beneficial in terms of risk and cost while keeping the widest safety margins possible.

  17. Development of HELIOS/CAPP code system for the analysis of block type VHTR cores

    SciTech Connect

    Lee, H. C.; Han, T. Y.; Jo, C. K.; Noh, J. M.

    2012-07-01

    In this paper, the HELIOS/CAPP code system developed for the analysis of block type VHTR cores is presented and verified against several VHTR core configurations. Verification results shows that HELIOS code predicts less negative MTC and RTC than McCARD code does and thus HELIOS code overestimates the multiplication factors at the states with high moderator and reflector temperature especially when the B{sub 4}C BP is loaded. In the depletion calculation for the VHTR single cell fuel element, the error of HELIOS code increases as burnup does. It is ascribed to the fact that HELIOS code treats some fission product nuclides with large resonances as non-resonant nuclides. In the 2-D core depletion calculation, a relatively large reactivity error is observed in the case with BP loading while the reactivity error in the case without BP loading is less than 300 pcm. (authors)

  18. Calculated Coupling Efficiency Between an Elliptical-Core Optical Fiber and a Silicon Oxynitride Rib Waveguide [Corrected Copy

    NASA Technical Reports Server (NTRS)

    Tuma, Margaret L.; Beheim, Glenn

    1995-01-01

    The effective-index method and Marcatili's technique were utilized independently to calculate the electric field profile of a rib channel waveguide. Using the electric field profile calculated from each method, the theoretical coupling efficiency between a single-mode optical fiber and a rib waveguide was calculated using the overlap integral. Perfect alignment was assumed and the coupling efficiency calculated. The coupling efficiency calculation was then repeated for a range of transverse offsets.

  19. Development and experimental validation of a calculation scheme for nuclear heating evaluation in the core of the OSIRIS material testing reactor

    SciTech Connect

    Malouch, F.

    2011-07-01

    The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

  20. Triton burnup study in JT-60U

    SciTech Connect

    Nishitani, T.; Hoek, M.; Isobe, M.; Tobita, K.; Kusama, Y.; Harano, H.; Wurden, G.A.; Chrien, R.E.

    1995-10-01

    The behavior of 1 MeV tritons produced in the d(d,p)t reaction is important to predict the properties of D-T produced 3.5 MeV alphas because 1 MeV tritons and 3.5 MeV alphas have similar kinematic properties, such as Larmor radius and precession frequency. The confinement and slowing down of the fast tritons were investigated by measuring the 14 MeV and the 2.5 MeV neutron production rates. Here the time resolved triton burnup measurements have been performed using a new type 14 MeV neutron detector based on scintillating fibers, as part of a US-Japan tokamak collaboration. Loss of alpha particles due to toroidal ripple is one of the most important issues to be solved for a fusion reactor such as ITER. The authors investigated the toroidal ripple effect on the fast triton by analyzing the time history of the 14 MeV emission after NB turn-off.

  1. Burn-up and neutron economy of accelerator-driven reactor

    SciTech Connect

    Takahashi, H.; Yang, W.; An, Y.; Yamazaki, Y.

    1997-07-01

    It is desirable to have only a small reactivity change in the large burn-up of a solid fuel fast reactor, so that the number of replacements or shuffling of the fuel can be reduced, and plant factor accordingly increased. Also, this reduces the number of control rods needed for the change in burn-up reactivity. In subcritical operation, power controlled by beam power is suggested, but this practice is not as economical as the use of control rods and makes more careful operation of the accelerator is required due to changes in the wake field. In subcritical operation, even a slightly subcritical one, the safety problems associated with a hard neutron spectrum can be alleviated. Neutron leakage from a flattened core, which is needed for operation of the critical fast reactor can be lessen by using the non flat core which has good neutron economy. For generating nuclear energy, it is essential to have a high neutron economy, although breeding the fuel is not welcomed in the present political climate, as is needed for transmuting long lived fission products. In contrast to the breeder, the accelerator driven reactor can separate the energy production from fuel production and processing. Thus, it is suited for non-proliferation of nuclear material by prohibiting the processing and production of fuel in the unrestricted area so this can be only done in international controlled areas which are restricted and remote.

  2. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

    SciTech Connect

    Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; Wagner, John C.

    2014-11-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

  3. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

    DOE PAGES

    Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; Wagner, John C.

    2014-11-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application ofmore » the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.« less

  4. Reactivity and isotopic composition of spent PWR (pressurized-water-reactor) fuel as a function of initial enrichment, burnup, and cooling time

    SciTech Connect

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub infinity/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub infinity/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub infinity/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs.

  5. Thermal Behavior of Advanced UO{sub 2} Fuel at High Burnup

    SciTech Connect

    Muller, E.; Lambert, T.; Silberstein, K.; Therache, B.

    2007-07-01

    To improve the fuel performance, advanced UO{sub 2} products are developed to reduce significantly Pellet-Cladding Interaction and Fission Gas Release to increase high burnup safety margins on Light Water Reactors. To achieve the expected improvements, doping elements are currently used, to produce large grain viscoplastic UO{sub 2} fuel microstructures. In that scope, AREVA NP is conducting the qualification of a new UO{sub 2} fuel pellet obtained by optimum chromium oxide doping. To assess the fuel thermal performance, especially the fuel conductivity degradation with increasing burnup and also the kinetics of fission gas release under transient operating conditions, an instrumented in-pile experiment, called REMORA, has been developed by the CEA. One segment base irradiated for five cycles in a French EDF commercial PWR ({approx} 62 GWd/tM) was consequently re-instrumented with a fuel centerline thermocouple and an advanced pressure sensor. The design of this specific sensor is based on the counter-pressure principle and avoids any drift phenomenon due to nuclear irradiation. This rodlet was then irradiated in the GRIFFONOS rig of the Osiris experimental reactor at CEA Saclay. This device, located in the periphery of the core, is designed to perform test under conditions close to those prevailing in French PWR reactor. Power variations are carried out by translating the device relatively to the core. Self - powered neutron detectors are positioned in the loop in order to monitor the power the whole time of the irradiation. The re-irradiation of the REMORA experiment consisted of a stepped ramp to power in order to point out a potential degradation of the fuel thermal conductivity with increasing burnup. During the first part of the irradiation, most of the measurements were performed at low power in order to take into account the irradiation effects on UO{sub 2} thermal conductivity at high burnup in low range of temperature. The second part of the irradiation

  6. Analysis of MNSR core composition changes using the codes WIMSD-4 and CITATION.

    PubMed

    Haj Hassan, H; Ghazi, N; Hainoun, A

    2008-10-01

    The codes WIMSD/4 and BORGES--part of the MTR-PC code package--have been applied to prepare the microscopic cross-section library for the main elements of miniature neutron source reactor (MNSR) core for six neutron energy groups. The generated library has been utilized by the 3D code CITATION to perform the calculation of fuel burn-up including the identification of main fission products and their impacts on the multiplication factor. In this regard some modifications have been introduced to the subroutine NUCY in CITATION to incorporate estimating the concentration of the related actinides and fission products. The burn-up results have indicated that the core life-time of MNSR is being mainly estimated by Sm(149) followed by Gd(157) and Cd(113). The accumulation of these fission products during 100 continuous operation days caused a reduction of about 4.3 mk for the excess reactivity. This result seems to be in good agreement with the available empirical value of 3.5 mk, which relates to the whole discontinuous operation period of the reactor since its start up to now. The calculation procedure simulates the sporadic operation with an equivalent continuous operation period. This approximation is valid for the long-lived fission products that mainly dictate the core life-time. However, it is an overestimation for the concentration of short-lived radioactive products like Xe(135).

  7. Modeling Cladding-Coolant Heat Transfer of High-Burnup Fuel During RIA

    SciTech Connect

    Wenfeng Liu; Kazimi, Mujid S.

    2006-07-01

    This paper describes a model for the cladding-coolant heat transfer of high burnup fuel during a Reactivity Initiated Accident (RIA) which is implemented in the fuel performance code FRAPTRAN 1.2. The minimum stable film boiling temperature, affected by the subcooling and the clad oxidation, is modeled by a modified Henry correlation. This accounts for the effects of thermal properties of the cladding surface on the transient temperature drop during liquid-solid contact. The transition boiling regime is described as the interpolation of the heat flux between two anchor points on the boiling curve: the Critical Heat Flux (CHF) and minimum stable film boiling. The CHF correlation is based on the Zuber hydrodynamic model multiplied by a subcooling factor. Frederking correlation is chosen to model the film boiling regime. The heat conduction through the oxide layer of the cladding surface of high burnup fuel is calculated by solving heat conduction equations with thermal properties of zirconia taken from MATPRO. This model is validated in the FRAPTRAN code for test cases of both high burnup and fresh test fuel rods including the burnup level (0--56 MW d/kg), peak fuel enthalpy deposit (70--190 cal/g), degree of subcooling (0--80 deg. C), and extent of oxidation (0--25 micron). The modified code demonstrates the capability of differentiating between the departure from nucleate boiling (DNB) and none-DNB cases. The predicted peak cladding temperature (PCT) and duration of DNB achieves generally good agreement with the experimental data. It is found that the cladding surface oxidation of high burnup fuel causes an early rewetting of cladding or suppresses DNB due to two factors: 1) Thick zirconia layer may delay the heat conducted to the surface while keeping the surface heat transfer in the most effective nucleate boiling regime. 2) The transient liquid-solid contact resulting from vapor breaking down would cause a lower interface temperature for an oxidized surface

  8. Plasmon-enhanced optical properties of Au/TiO2 core-shell nanowires studied by finite difference time domain calculation

    NASA Astrophysics Data System (ADS)

    Lee, Jubok; Lee, Sun-Hee; Kim, Min Su; Shin, Hyungjung; Kim, Jeongyong

    2014-09-01

    We performed Finite Difference Time Domain (FDTD) calculation to investigate the enhancement of optical properties such as light scattering and absorption of Au-hybridized TiO2 core-shell structures which can lead to the improvement of photocatalytic and solar cell performance. The results showed that by hybridization of Au as core and TiO2 as shell provides the significant enhancement of light scattering and absorption. Furthermore, the tuning of scattering resonance wavelength may be achieved by varying the diameter of Au core. Our result suggests that hybridization Au and TiO2, with proper introduction of interband states in TiO2, can increase and color-tune the photocatalytic efficiency and solar cell performance of TiO2 nanostructures.

  9. PWR cores with silicon carbide cladding

    SciTech Connect

    Dobisesky, J. P.; Carpenter, D.; Pilat, E.; Kazimi, M. S.

    2012-07-01

    The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

  10. Spent fuel burnup estimation by Cerenkov glow intensity measurement

    SciTech Connect

    Kuribara, Masayuki . Communication and Information Research Lab.)

    1994-10-01

    The Cerenkov glow images from irradiated fuel assemblies of boiling-water reactors (BWR) and pressurized-water reactors (PWR) are generally used for inspections. For this purpose, a new UV-I.I. CVD (ultra-violet light image intensifier Cerenkov viewing device), has been developed. This new device can measure the intensity of the Cerenkov glow from a spent fuel assembly, thus making it possible to estimate the burnup of the fuel assembly by comparing the Cerenkov glow intensity to the reference intensity. The experiment was carried out on BWR spent fuel assemblies and the results show that burnups are estimated within 20% accuracy compared to the declared burnups for the tested spent fuel assemblies for cooling times ranging from 900--2.000 d.

  11. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    SciTech Connect

    Pecchia, M.; D'Auria, F.; Mazzantini, O.

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  12. Isotopic biases for actinide-only burnup credit

    SciTech Connect

    Rahimi, M.; Lancaster, D.; Hoeffer, B.; Nichols, M.

    1997-04-01

    The primary purpose of this paper is to present the new methodology for establishing bias and uncertainty associated with isotopic prediction in spent fuel assemblies for burnup credit analysis. The analysis applies to the design of criticality control systems for spent fuel casks. A total of 54 spent fuel samples were modeled and analyzed using the Shielding Analyses Sequence (SAS2H). Multiple regression analysis and a trending test were performed to develop isotopic correction factors for 10 actinide burnup credit isotopes. 5 refs., 1 tab.

  13. Isotopic and criticality validation for actinide-only burnup credit

    SciTech Connect

    Fuentes, E.; Lancaster, D.; Rahimi, M.

    1997-07-01

    The techniques used for actinide-only burnup credit isotopic validation and criticality validation are presented and discussed. Trending analyses have been incorporated into both methodologies, requiring biases and uncertainties to be treated as a function of the trending parameters. The isotopic validation is demonstrated using the SAS2H module of SCALE 4.2, with the 27BURNUPLIB cross section library; correction factors are presented for each of the actinides in the burnup credit methodology. For the criticality validation, the demonstration is performed with the CSAS module of SCALE 4.2 and the 27BURNUPLIB, resulting in a validated upper safety limit.

  14. Efficiency and accuracy of the perturbation response coefficient generation method for whole core comet calculations in BWR and CANDU configurations

    SciTech Connect

    Zhang, D.; Rahnema, F.

    2013-07-01

    The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)

  15. Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations.

    1994-11-15

    Version 00 The MARIA System calculates cross sections for PWR fuel assembly calculations. It generates the cross sections library for the diffusion calculations with burnup and feedback effects (CARMEN System, NEA 0649 and RSIC CCC-487) and the k(infinite) and M**2 parameters for the nodal calculations (SIMULA, NEA 0768). MARIA includes three modules. PRELIM generates the input data for the fuel assembly calculation module, for all fuel assembly types in the core and at any conditionmore » of power rate and temperature. WIMS-TRACA is a modified version of the fuel assembly calculation program WIMS-D/4 (NEA 0329 and RSIC CCC-576), which generates the collapsed cross sections versus burn up needed by the CARMEN code (reference cell, boron, xenon, samarium, and light water). POSWIM calculates the transport corrections to the diffusion constant of the absorber materials generated by WIMS-TRACA, to be used directly in the diffusion code when rods or burnable absorber rods are present.« less

  16. PAMELA: An open-source software package for calculating nonlocal exact exchange effects on electron gases in core-shell nanowires

    NASA Astrophysics Data System (ADS)

    Long, Andrew W.; Wong, Bryan M.

    2012-09-01

    We present a new pseudospectral approach for incorporating many-body, nonlocal exact exchange interactions to understand the formation of electron gases in core-shell nanowires. Our approach is efficiently implemented in the open-source software package PAMELA (Pseudospectral Analysis Method with Exchange & Local Approximations) that can calculate electronic energies, densities, wavefunctions, and band-bending diagrams within a self-consistent Schrödinger-Poisson formalism. The implementation of both local and nonlocal electronic effects using pseudospectral methods is key to PAMELA's efficiency, resulting in significantly reduced computational effort compared to finite-element methods. In contrast to the new nonlocal exchange formalism implemented in this work, we find that the simple, conventional Schrödinger-Poisson approaches commonly used in the literature (1) considerably overestimate the number of occupied electron levels, (2) overdelocalize electrons in nanowires, and (3) significantly underestimate the relative energy separation between electronic subbands. In addition, we perform several calculations in the high-doping regime that show a critical tunneling depth exists in these nanosystems where tunneling from the core-shell interface to the nanowire edge becomes the dominant mechanism of electron gas formation. Finally, in order to present a general-purpose set of tools that both experimentalists and theorists can easily use to predict electron gas formation in core-shell nanowires, we document and provide our efficient and user-friendly PAMELA source code that is freely available at http://alum.mit.edu/www/usagi.

  17. Applications of a Monte Carlo whole-core microscopic depletion method

    SciTech Connect

    Hutton, J.L.; Butement, A.W.; Watt, S.; Shadbolt, R.D.

    1995-12-31

    In the WIMS-6 (Ref. 1) reactor physics program scheme a three-dimensional microscopic depletion method has been developed using Monte Carlo fluxes. Together with microscopic cross sections, these give nuclide reaction rates, which are used to solve nuclide depletion equations for each region. An extension of the method, enabling rapid whole-core calculations, has been implemented in the long-established companion code MONK5W. Predictions at successive depletion time steps are based on a calculational route where both geometry and cross sections are accurately represented, providing a reliable and independent approach for benchmarking other methods. Newly developed tracking and storage procedures in MONK5W enable whole core burnup modeling on a desktop computer. Theory and applications are presented in this paper.

  18. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions

    SciTech Connect

    Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina; Wagner, John C

    2011-01-01

    The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for

  19. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    SciTech Connect

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

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

    SciTech Connect

    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.

  1. Vortices of polarization in BaTiO3 core-shell nanoceramics: Calculations based on ab initio derived Hamiltonian versus Landau theory

    NASA Astrophysics Data System (ADS)

    Anoufa, M.; Kiat, J. M.; Kornev, I.; Bogicevic, C.

    2013-10-01

    In this paper, we want to emphasize the fact that many experimental properties of ceramics can be explained by the existence of a core-shell structure of the grains, particularly at small sizes. In this framework, we have studied BaTiO3 (BT) ceramics constituted of core-shell nanoparticles, nanowires, or nanoplanes by using ab initio derived effective Hamiltonian calculations whose application range is for large values of shell thickness and low values of shell permittivity. Many differences and new features compared to the situation of nanodots are induced by the core-shell structure. For instance, phase sequences are different; there is also a coexistence of vortices found by Naumov, Bellaiche, and Fu [I. I. Naumov, L. Bellaiche, and H. Fu, Nature (London)10.1038/nature03107 432, 737 (2004)] in the case of isolated dots with a homogeneous polarization, a transition from cubic paraelectric phase towards nonpolar rhombohedral phase, anomalies in dielectric permittivity associated with the onset of toroidal moments, etc. Afterwards, we compare these results with those obtained by the Landau theory of core-shell ceramics we have recently published. However, the ab initio calculations fail to capture the physics at small shell thickness and/or high shell permittivity, whereas the Landau theory fails to predict the peculiar properties of the phases in which vortices exist. Therefore, in a tentative way to build a global theory, we have constructed a Landau potential using both the polarization and the toroidal moment as competing order parameters, which allows us to propose a phase diagram, whatever the thickness and permittivity of the shell are.

  2. First-principles core-level X-ray photoelectron spectroscopy calculation on arsenic defects in silicon crystal

    SciTech Connect

    Kishi, Hiroki; Miyazawa, Miki; Matsushima, Naoki; Yamauchi, Jun

    2014-02-21

    We investigate the X-ray photoelectron spectroscopy (XPS) binding energies of As 3d in Si for various defects in neutral and charged states by first-principles calculation. It is found that the complexes of a substitutional As and a vacancy in charged and neutral states explain the experimentally observed unknown peak very well.

  3. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    SciTech Connect

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.

  4. Towards a spectroscopically accurate set of potentials for heavy hydride laser cooling candidates: Effective core potential calculations of BaH

    NASA Astrophysics Data System (ADS)

    Moore, Keith; McLaughlin, Brendan M.; Lane, Ian C.

    2016-04-01

    BaH (and its isotopomers) is an attractive molecular candidate for laser cooling to ultracold temperatures and a potential precursor for the production of ultracold gases of hydrogen and deuterium. The theoretical challenge is to simulate the laser cooling cycle as reliably as possible and this paper addresses the generation of a highly accurate ab initio 2Σ+ potential for such studies. The performance of various basis sets within the multi-reference configuration-interaction (MRCI) approximation with the Davidson correction is tested and taken to the Complete Basis Set (CBS) limit. It is shown that the calculated molecular constants using a 46 electron effective core-potential and even-tempered augmented polarized core-valence basis sets (aug-pCVnZ-PP, n = 4 and 5) but only including three active electrons in the MRCI calculation are in excellent agreement with the available experimental values. The predicted dissociation energy De for the X2Σ+ state (extrapolated to the CBS limit) is 16 895.12 cm-1 (2.094 eV), which agrees within 0.1% of a revised experimental value of <16 910.6 cm-1, while the calculated re is within 0.03 pm of the experimental result.

  5. Need for higher fuel burnup at the Hatch Plant

    SciTech Connect

    Beckhman, J.T.

    1996-03-01

    Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make the conversions if you think in one way or the other because I am talking in batch averages. During Hatch`s operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about.

  6. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR using a U7Mo Dispersion Fuel

    SciTech Connect

    Hanson A. L.; Diamond D.

    2014-06-30

    A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The LEU fuel may be a monolithic foil (LEUm) of U10Mo (10% molybdenum by weight in an alloy with uranium) or a dispersion of U7Mo in aluminum (LEUd). A previous report provided neutronic calculations for the LEUm fuel and this report presents the neutronics parameters for the LEUd fuel. The neutronics parameters for the LEUd fuel are compared to those previously obtained for the present HEU fuel and the proposed LEUm fuel. The results show no significant differences between the LEUm and the LEUd other than the LEUd fuel requires slightly less uranium than the LEUm fuel due to less molybdenum being present. The calculations include kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions under normal operation and with misloaded fuel elements.

  7. Relativistic many-body calculations of excitation energies and transition rates from core-excited states in copperlike ions

    NASA Astrophysics Data System (ADS)

    Safronova, U. I.; Johnson, W. R.; Shlyaptseva, A.; Hamasha, S.

    2003-05-01

    Energies of (3s23p63d94l4l'), (3s23p53d104l4l'), and (3s3p63d104l4l') states for Cu-like ions with Z=30 100 are evaluated to second order in relativistic many-body perturbation theory (RMBPT) starting from a Ni-like Dirac-Fock potential. Second-order Coulomb and Breit-Coulomb interactions are included. Correction for the frequency dependence of the Breit interaction is taken into account in lowest order. The Lamb shift correction to energies is also included in lowest order. Intrinsic particle-particle-hole contributions to energies are found to be 20 30 % of the sum of one- and two-body contributions. Transition rates and line strengths are calculated for the 3l-4l' electric-dipole (E1) transitions in Cu-like ions with nuclear charge Z=30 100. RMBPT including the Breit interaction is used to evaluate retarded E1 matrix elements in length and velocity forms. First-order RMBPT is used to obtain intermediate coupling coefficients, and second-order RMBPT is used to calculate transition matrix elements. A detailed discussion of the various contributions to the dipole- matrix elements and energy levels is given for copperlike tungsten (Z=74). The transition energies used in the calculation of oscillator strengths and transition rates are from second-order RMBPT. Trends of the transition rates as functions of Z are illustrated graphically for selected transitions. Comparisons are made with available experimental data. These atomic data are important in the modeling of M-shell radiation spectra of heavy ions generated in electron-beam ion trap experiments and in M-shell diagnostics of plasmas.

  8. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    NASA Astrophysics Data System (ADS)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  9. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    SciTech Connect

    Widiawati, Nina Su’ud, Zaki

    2015-09-30

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from −0.6695443 % at BOC to −0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  10. NICIL: A Stand Alone Library to Self-Consistently Calculate Non-Ideal Magnetohydrodynamic Coefficients in Molecular Cloud Cores

    NASA Astrophysics Data System (ADS)

    Wurster, James

    2016-09-01

    In this paper, we introduce Nicil: Non-Ideal magnetohydrodynamics Coefficients and Ionisation Library. Nicil is a stand-alone Fortran90 module that calculates the ionisation values and the coefficients of the non-ideal magnetohydrodynamics terms of Ohmic resistivity, the Hall effect, and ambipolar diffusion. The module is fully parameterised such that the user can decide which processes to include and decide upon the values of the free parameters, making this a versatile and customisable code. The module includes both cosmic ray and thermal ionisation; the former includes two ion species and three species of dust grains (positively charged, negatively charged, and neutral), and the latter includes five elements which can be doubly ionised. We demonstrate tests of the module, and then describe how to implement it into an existing numerical code.

  11. Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor.

    PubMed

    Dawahra, S; Khattab, K; Saba, G

    2015-07-01

    Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of (135)Xe and (149)Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the (135)Xe and (149)Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The (149)Sm reactivities could be neglected compared to (135)Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4-Al produced the highest (135)Xe reactivity in all the studied fuel group during the reactor operation (0.39 mk) and after the reactor shutdown (0.735 mk), It followed by U3Si-Al (0.34 mk, 0.653 mk), U3Si2-Al (0.33 mk, 0.634 mk), U9Mo-Al (0.3 mk, 0.568 mk) and UO2 (0.24 mk, 0.448 mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest (135)Xe reactivity during the reactor operation and after shutdown.

  12. Design and analysis of a nuclear reactor core for innovative small light water reactors

    NASA Astrophysics Data System (ADS)

    Soldatov, Alexey I.

    In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

  13. Study on Ultra-Long Life,Small U-Zr Metallic Fuelled Core With Burnable Poison

    SciTech Connect

    Kenji Tsuji; Hiromitsu Inagaki; Akira Nishikawa; Hisato Matsumiya; Yoshiaki Sakashita; Yasuyuki Moriki; Mitsuaki Yamaoka; Norihiko Handa

    2002-07-01

    A conceptual design for a 50 MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled core has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up. (authors)

  14. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

    burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

  15. The Fork+ burnup measurement system: Design and first measurement campaign

    SciTech Connect

    Olson, C.E.; Bronowski, D.R.; McMurtry, W.; Ewing, R.; Jordan, R.; Rivard, D.

    1998-12-31

    Previous work with the original Fork detector showed that burnup as determined by reactor records could be accurately allocated to spent nuclear fuel assemblies. The original Fork detector, designed by Los Alamos National Laboratory, used an ion chamber to measure gross gamma count and a fission chamber to measure neutrons from an activation source, {sup 244}Cm. In its review of the draft Topical Report on Burnup Credit, the US Nuclear Regulatory Commission indicated it felt uncomfortable with a measurement system that depended on reactor records for calibration. The Fork+ system was developed at Sandia National Laboratories under the sponsorship of the Electric Power Research Institute with the aim of providing this independent measurement capability. The initial Fork+ prototype was used in a measurement campaign at the Maine Yankee reactor. The campaign confirmed the applicability of the sensor approach in the Fork+ system and the efficiency of the hand-portable Fork+ prototype in making fuel assembly measurements. It also indicated potential design modifications that will be necessary before the Fork+ can be used effectively on high-burnup spent fuel.

  16. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    NASA Astrophysics Data System (ADS)

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

    2014-06-01

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  17. The effect of coolant orificing on the core performance of a heterogeneous liquid-metal fast breeder reactor

    SciTech Connect

    Mamoru, K.; Shigehiro, A.; Yoshiaki, O.

    1983-04-01

    The effect of orificing on the core performance of a commercial-size heterogeneous liquid-metal fast breeder reactor was studied analytically. The thermal power output was flattened at beginning of life, and the coolant flow rate was chosen such that the maximum inner cladding temperature of a driver fuel and a blanket fuel was less than or equal to 620/sup 0/C at both beginning of equilibrium life (BOEL) and end of equilibrium life (EOEL). The difference between reactor outlet temperatures at BOEL and EOEL was then calculated for six core configurations: one homogeneous core configuration and five heterogeneous ones. The results showed that the core outlet temperature variation due to the change of the power profile of the radial heterogeneous core configurations is similar to that of the homogeneous one, even when a single type of orificing is used in each core zone, and it will not be necessary to use the more detailed orificing in each zone of a heterogeneous core configuration. The study concludes that for the present design, especially the thermal design, of some heterogeneous core configurations, it is feasible to control the change of the reactor outlet temperature with burnup, even when a single type of orificing is used in each core zone.

  18. Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

    SciTech Connect

    Leal, Luiz C; Derrien, Herve; Dunn, Michael E; Mueller, Don

    2007-12-01

    Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the

  19. Extended burnup fuel - a beneficial environmental step for the nuclear fuel cycle

    SciTech Connect

    Tulenko, J.S. )

    1992-01-01

    The beneficial environmental effects generated by extending the burnup of light water reactors (LWR) fuel covers the entire fuel cycle from cradle (mining) to grave (fuel disposal). The purpose of this paper is to outline the beneficial efforts of extending burnup and show how a proposed financial incentive waste charge would encourage the development and utilization of extended burnup. Environmental consciousness demands that economic incentives be implemented to encourage utilities to extend burnup of their LWR fuel. This paper builds on earlier work and focuses and extends their results. The cornerstone of the positive environmental effects of extended burnup results from the increased plutonium energy generation as burnup is increased and residence time is increased.

  20. Effects of microstructural constraints on the transport of fission products in uranium dioxide at low burnups

    NASA Astrophysics Data System (ADS)

    Lim, Harn Chyi; Rudman, Karin; Krishnan, Kapil; McDonald, Robert; Dickerson, Patricia; Gong, Bowen; Peralta, Pedro

    2016-08-01

    Diffusion of fission gases in UO2 is studied at low burnups, before bubble growth and coalescence along grain boundaries (GBs) become dominant, using a 3-D finite element model that incorporates actual UO2 microstructures. Grain boundary diffusivities are assigned based on crystallography with lattice and GB diffusion coupled with temperature to account for temperature gradients. Heterogeneity of GB properties and connectivity can induce regions where concentration is locally higher than without GB diffusion. These regions are produced by "bottlenecks" in the GB network because of lack of connectivity among high diffusivity GBs due to crystallographic constraints, and they can lead to localized swelling. Effective diffusivities were calculated assuming a uniform distribution of high diffusivity among GBs. Results indicate an increase over the bulk diffusivity with a clear grain size effect and that connectivity and properties of different GBs become important factors on the variability of fission product concentration at the microscale.

  1. Determination of nuclear fuel burn-up axial profile by neutron emission measurement

    NASA Astrophysics Data System (ADS)

    Prokopowicz, Rafal; Pytel, Krzysztof

    2016-12-01

    Burning-up of nuclear fuel is usually not a space-isotropic phenomenon. It depends on both the neutron flux density and energy spectrum distribution during fuel operation in a nuclear reactor. This paper presents the method of measurement of burn-up spatial distribution of spent nuclear fuel element. The method is based on recording of the neutron emission from investigated fuel element. Based on performed analyses and calculations, a suitable measuring setup has been designed and constructed. The subjects of investigation were fuel elements used in the MARIA research reactor, operated by National Centre for Nuclear Research in Świerk, Poland. The results of measurements made over a period of several years by means of the described method are presented in the paper.

  2. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    NASA Astrophysics Data System (ADS)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  3. Using SERPENT Monte Carlo and Burnup code to model Traveling Wave Reactors (TWR)

    NASA Astrophysics Data System (ADS)

    Gulik, Volodymyr; Pavlovych, Volodymyr; Tkaczyk, Alan Henry

    2014-06-01

    This paper is mainly devoted to the proof-of-principle implementation of the SERPENT code for the simulation of traveling wave reactors. The investigation of SERPENT 1.1.19 code for multiprocessor tasks with long burnup steps was performed. The investigation of SERPENT 2 code for multiprocessor tasks with long burnup steps was performed. Methods to remove the influence of the ignition zone were considered, and neutronics simulations with various fragmentations of the burnup zone were performed.

  4. Gamma-Ray Simulated Spectrum Deconvolution of a LaBr₃ 1-in. x 1-in. Scintillator for Nondestructive ATR Fuel Burnup On-Site Predictions

    DOE PAGES

    Navarro, Jorge; Ring, Terry A.; Nigg, David W.

    2015-03-01

    A deconvolution method for a LaBr₃ 1"x1" detector for nondestructive Advanced Test Reactor (ATR) fuel burnup applications was developed. The method consisted of obtaining the detector response function, applying a deconvolution algorithm to 1”x1” LaBr₃ simulated, data along with evaluating the effects that deconvolution have on nondestructively determining ATR fuel burnup. The simulated response function of the detector was obtained using MCNPX as well with experimental data. The Maximum-Likelihood Expectation Maximization (MLEM) deconvolution algorithm was selected to enhance one-isotope source-simulated and fuel- simulated spectra. The final evaluation of the study consisted of measuring the performance of the fuel burnup calibrationmore » curve for the convoluted and deconvoluted cases. The methodology was developed in order to help design a reliable, high resolution, rugged and robust detection system for the ATR fuel canal capable of collecting high performance data for model validation, along with a system that can calculate burnup and using experimental scintillator detector data.« less

  5. Gamma-Ray Simulated Spectrum Deconvolution of a LaBr₃ 1-in. x 1-in. Scintillator for Nondestructive ATR Fuel Burnup On-Site Predictions

    SciTech Connect

    Navarro, Jorge; Ring, Terry A.; Nigg, David W.

    2015-03-01

    A deconvolution method for a LaBr₃ 1"x1" detector for nondestructive Advanced Test Reactor (ATR) fuel burnup applications was developed. The method consisted of obtaining the detector response function, applying a deconvolution algorithm to 1”x1” LaBr₃ simulated, data along with evaluating the effects that deconvolution have on nondestructively determining ATR fuel burnup. The simulated response function of the detector was obtained using MCNPX as well with experimental data. The Maximum-Likelihood Expectation Maximization (MLEM) deconvolution algorithm was selected to enhance one-isotope source-simulated and fuel- simulated spectra. The final evaluation of the study consisted of measuring the performance of the fuel burnup calibration curve for the convoluted and deconvoluted cases. The methodology was developed in order to help design a reliable, high resolution, rugged and robust detection system for the ATR fuel canal capable of collecting high performance data for model validation, along with a system that can calculate burnup and using experimental scintillator detector data.

  6. Two-component relativistic density-functional calculations of the dimers of the halogens from bromine through element 117 using effective core potential and all-electron methods.

    PubMed

    Mitin, Alexander V; van Wüllen, Christoph

    2006-02-14

    A two-component quasirelativistic Hamiltonian based on spin-dependent effective core potentials is used to calculate ionization energies and electron affinities of the heavy halogen atom bromine through the superheavy element 117 (eka-astatine) as well as spectroscopic constants of the homonuclear dimers of these atoms. We describe a two-component Hartree-Fock and density-functional program that treats spin-orbit coupling self-consistently within the orbital optimization procedure. A comparison with results from high-order Douglas-Kroll calculations--for the superheavy systems also with zeroth-order regular approximation and four-component Dirac results--demonstrates the validity of the pseudopotential approximation. The density-functional (but not the Hartree-Fock) results show very satisfactory agreement with theoretical coupled cluster as well as experimental data where available, such that the theoretical results can serve as an estimate for the hitherto unknown properties of astatine, element 117, and their dimers. PMID:16483205

  7. Accelerated event-by-event Monte Carlo microdosimetric calculations of electrons and protons tracks on a multi-core CPU and a CUDA-enabled GPU.

    PubMed

    Kalantzis, Georgios; Tachibana, Hidenobu

    2014-01-01

    For microdosimetric calculations event-by-event Monte Carlo (MC) methods are considered the most accurate. The main shortcoming of those methods is the extensive requirement for computational time. In this work we present an event-by-event MC code of low projectile energy electron and proton tracks for accelerated microdosimetric MC simulations on a graphic processing unit (GPU). Additionally, a hybrid implementation scheme was realized by employing OpenMP and CUDA in such a way that both GPU and multi-core CPU were utilized simultaneously. The two implementation schemes have been tested and compared with the sequential single threaded MC code on the CPU. Performance comparison was established on the speed-up for a set of benchmarking cases of electron and proton tracks. A maximum speedup of 67.2 was achieved for the GPU-based MC code, while a further improvement of the speedup up to 20% was achieved for the hybrid approach. The results indicate the capability of our CPU-GPU implementation for accelerated MC microdosimetric calculations of both electron and proton tracks without loss of accuracy.

  8. First-principle calculation of core level binding energies of Li{sub x}PO{sub y}N{sub z} solid electrolyte

    SciTech Connect

    Guille, Émilie; Vallverdu, Germain Baraille, Isabelle

    2014-12-28

    We present first-principle calculations of core-level binding energies for the study of insulating, bulk phase, compounds, based on the Slater-Janak transition state model. Those calculations were performed in order to find a reliable model of the amorphous Li{sub x}PO{sub y}N{sub z} solid electrolyte which is able to reproduce its electronic properties gathered from X-ray photoemission spectroscopy (XPS) experiments. As a starting point, Li{sub 2}PO{sub 2}N models were investigated. These models, proposed by Du et al. on the basis of thermodynamics and vibrational properties, were the first structural models of Li{sub x}PO{sub y}N{sub z}. Thanks to chemical and structural modifications applied to Li{sub 2}PO{sub 2}N structures, which allow to demonstrate the relevance of our computational approach, we raise an issue concerning the possibility of encountering a non-bridging kind of nitrogen atoms (=N{sup −}) in Li{sub x}PO{sub y}N{sub z} compounds.

  9. Global variance reduction for Monte Carlo reactor physics calculations

    SciTech Connect

    Zhang, Q.; Abdel-Khalik, H. S.

    2013-07-01

    Over the past few decades, hybrid Monte-Carlo-Deterministic (MC-DT) techniques have been mostly focusing on the development of techniques primarily with shielding applications in mind, i.e. problems featuring a limited number of responses. This paper focuses on the application of a new hybrid MC-DT technique: the SUBSPACE method, for reactor analysis calculation. The SUBSPACE method is designed to overcome the lack of efficiency that hampers the application of MC methods in routine analysis calculations on the assembly level where typically one needs to execute the flux solver in the order of 10{sup 3}-10{sup 5} times. It places high premium on attaining high computational efficiency for reactor analysis application by identifying and capitalizing on the existing correlations between responses of interest. This paper places particular emphasis on using the SUBSPACE method for preparing homogenized few-group cross section sets on the assembly level for subsequent use in full-core diffusion calculations. A BWR assembly model is employed to calculate homogenized few-group cross sections for different burn-up steps. It is found that using the SUBSPACE method significant speedup can be achieved over the state of the art FW-CADIS method. While the presented speed-up alone is not sufficient to render the MC method competitive with the DT method, we believe this work will become a major step on the way of leveraging the accuracy of MC calculations for assembly calculations. (authors)

  10. Calculation of fuel pin failure timing under LOCA conditions

    SciTech Connect

    Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

    1991-10-01

    The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs.

  11. Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm

    SciTech Connect

    Soewono, C. N.; Takaki, N.

    2012-07-01

    In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

  12. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    NASA Astrophysics Data System (ADS)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  13. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    SciTech Connect

    Su'ud, Zaki; Sekimoto, H.

    2014-09-30

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  14. Model design on calculations of microwave permeability and permittivity of Fe/SiO2 particles with core/shell structure

    NASA Astrophysics Data System (ADS)

    Yang, Y.; Li, Z. W.; Neo, C. P.; Ding, J.

    2014-02-01

    Fe/SiO2 particles with core/shell structure were prepared by coating silica on the surface of a commercial spherical carbonyl iron via the hydrolysis process of tetraethyl orthosilicate (TEOS). The electromagnetic performance of commercial carbonyl iron and as-prepared Fe/SiO2 particles was studied theoretically and experimentally. As predicted by the theoretical calculation based on the Bruggeman formula and the Landau-Lifshitz-Gilbert (LLG) theory, the insulating surface layer of silica was effective to reduce the permittivity parameters of pure carbonyl iron. The measured results showed good agreement with the theoretical prediction. Although there was a little decrease in the permeability of the Fe/SiO2 core/shell particles, a better impedance match especially at higher frequency range was obtained when used as a microwave absorber. The reflection loss (RL) curves show that the lowest reflection loss of Fe/Epoxy composite (-20.5 GHz) was obtained corresponding to the frequency of 8.5 GHz when the thickness of the absorber was 3 mm. A different trend was observed in Fe/SiO2/Epoxy composite. The reflection loss value got lower by decreasing the thickness of absorbers. At the thickness of 2.2 mm, a relative low reflection loss (-17 GHz) corresponding to the frequency of 13.6 GHz was obtained. Compared with the Fe/Epoxy composite, the improvement on shifting the reflection loss peak to higher frequency and on reducing the optimal thickness of absorbers was made by Fe/SiO2/Epoxy composite.

  15. New results from the NSRR experiments with high burnup fuel

    SciTech Connect

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  16. Core physics analysis of 100% MOX Core in IRIS

    SciTech Connect

    Franceschini, F.; Petrovic, B.

    2006-07-01

    International Reactor Innovative and Secure (IRIS) is an advanced small-to-medium-size (1000 MWt) Pressurized Water Reactor (PWR), targeting deployment around 2015. Its reference core design is based on the current Westinghouse UO{sub 2} fuel with less than 5% {sup 235}U, and the analysis has been previously completed confirming good performance. The full MOX fuel core is currently under evaluation as one of the alternatives for the second wave of IRIS reactors. A full 3-D neutronic analysis has been performed to examine main core performance parameters, such as critical boron concentration, peaking factors, discharge burnup, etc. The enhanced moderation of the IRIS fuel lattice facilitates MOX core design, and all the obtained results are within the requirements, confirming viability of this option from the reactor physics standpoint. (authors)

  17. Attainable Burnup in a LIFE Engine Loaded with Depleted Uranium

    SciTech Connect

    Fratoni, M; Kramer, K J; Latkowski, J F

    2009-11-30

    The Laser Inertial Fusion-based Energy (LIFE) system uses a laser-based fusion source for electricity production. The (D,T) reaction, beside a pure fusion system, allows the option to drive a sub-critical fission blanket in order to increase the total energy gain. In a typical fusion-fission LIFE engine the fission blanket is a spherical shell around the fusion source, preceded by a beryllium shell for neutron multiplications by means of (n,2n) reactions. The fuel is in the form of TRISO particles dispersed in carbon pebbles, cooled by flibe. The optimal design features 80 cm thick blanket, 16 cm multiplier, and 20% TRISO packing factor. A blanket loaded with depleted uranium and depleted in a single batch with continuous mixing can achieve burnup as high as {approx}85% FIMA while generating 2,000 MW of total thermal power and producing enough tritium to be used for fusion. A multi-segment blanket with a central promotion shuffling scheme enhances burnup to {approx}90% FIMA, whereas a blanket that is operated with continuous refueling achieves only 82% FIMA under the same constraints of thermal power and tritium self-sufficiency. Both, multi-segment and continuous refueling eliminate the need for a fissile breeding phase.

  18. Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

    SciTech Connect

    Hikaru Hiruta; Gilles Youinou

    2013-09-01

    This report presents FY13 activities for the analysis of D2O cooled tight-pitch High-Conversion PWRs (HCPWRs) with U-Pu and Th-U fueled cores aiming at break-even or near breeder conditions while retaining the negative void reactivity. The analyses are carried out from several aspects which could not be covered in FY12 activities. SCALE 6.1 code system is utilized, and a series of simple 3D fuel pin-cell models are developed in order to perform Monte Carlo based criticality and burnup calculations. The performance of U-Pu fueled cores with axial and internal blankets is analyzed in terms of their impact on the relative fissile Pu mass balance, initial Pu enrichment, and void coefficient. In FY12, Pu conversion performances of D2O-cooled HCPWRs fueled with MOX were evaluated with small sized axial/internal DU blankets (approximately 4cm of axial length) in order to ensure the negative void reactivity, which evidently limits the conversion performance of HCPWRs. In this fiscal year report, the axial sizes of DU blankets are extended up to 30 cm in order to evaluate the amount of DU necessary to reach break-even and/or breeding conditions. Several attempts are made in order to attain the milestone of the HCPWR designs (i.e., break-even condition and negative void reactivity) by modeling of HCPWRs under different conditions such as boiling of D2O coolant, MOX with different 235U enrichment, and different target burnups. A similar set of analyses are performed for Th-U fueled cores. Several promising characteristics of 233U over other fissile like 239Pu and 235U, most notably its higher fission neutrons per absorption in thermal and epithermal ranges combined with lower ___ in the fast range than 239Pu allows Th-U cores to be taller than MOX ones. Such an advantage results in 4% higher relative fissile mass balance than that of U-Pu fueled cores while retaining the negative void reactivity until the target burnup of 51 GWd/t. Several other distinctions between U-Pu and

  19. Core Design Applications

    1995-07-12

    CORD-2 is intended for core desigh applications of pressurized water reactors. The main objective was to assemble a core design system which could be used for simple calculations (such as frequently required for fuel management) as well as for accurate calculations (for example, core design after refueling).

  20. Shutdown-induced tensile stress in monolithic miniplates as a possible cause of plate pillowing at very high burnup

    SciTech Connect

    Medvedev, Pavel G; Ozaltun, Hakan; Robinson, Adam Brady; Rabin, Barry H

    2014-04-01

    Post-irradiation examination of Reduced Enrichment for Research and Test Reactors (RERTR)-12 miniplates showed that in-reactor pillowing occurred in at least 4 plates, rendering performance of these plates unacceptable. To address in-reactor failures, efforts are underway to define the mechanisms responsible for in-reactor pillowing, and to suggest improvements to the fuel plate design and operational conditions. To achieve these objectives, the mechanical response of monolithic fuel to fission and thermally-induced stresses was modeled using a commercial finite element analysis code. Calculations of stresses and deformations in monolithic miniplates during irradiation and after the shutdown revealed that the tensile stress generated in the fuel increased from 2 MPa to 100 MPa at shutdown. The increase in tensile stress at shutdown possibly explains in-reactor pillowing of several RERTR-12 miniplates irradiated to the peak local burnup of up to 1.11x1022 fissions/cm3 . This paper presents the modeling approach and calculation results, and compares results with post-irradiation examinations and mechanical testing of irradiated fuel. The implications for the safe use of the monolithic fuel in research reactors are discussed, including the influence of fuel burnup and power on the magnitude of the shutdown-induced tensile stress.

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

    SciTech Connect

    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

  2. A multi-group Monte Carlo core analysis method and its application in SCWR design

    SciTech Connect

    Zhang, P.; Wang, K.; Yu, G.

    2012-07-01

    Complex geometry and spectrum have been the characteristics of many newly developed nuclear energy systems, so the suitability and precision of the traditional deterministic codes are doubtable while being applied to simulate these systems. On the contrary, the Monte Carlo method has the inherent advantages of dealing with complex geometry and spectrum. The main disadvantage of Monte Carlo method is that it takes long time to get reliable results, so the efficiency is too low for the ordinary core designs. A new Monte Carlo core analysis scheme is developed, aimed to increase the calculation efficiency. It is finished in two steps: Firstly, the assembly level simulation is performed by continuous energy Monte Carlo method, which is suitable for any geometry and spectrum configuration, and the assembly multi-group constants are tallied at the same time; Secondly, the core level calculation is performed by multi-group Monte Carlo method, using the assembly group constants generated in the first step. Compared with the heterogeneous Monte Carlo calculations of the whole core, this two-step scheme is more efficient, and the precision is acceptable for the preliminary analysis of novel nuclear systems. Using this core analysis scheme, a SCWR core was designed based on a new SCWR assembly design. The core output is about 1,100 MWe, and a cycle length of about 550 EFPDs can be achieved with 3-batch refueling pattern. The average and maximum discharge burn-up are about 53.5 and 60.9 MWD/kgU respectively. (authors)

  3. Analyse de l'impact de l'environnement dans un schema de calcul a deux etapes avec DRAGON et DONJON

    NASA Astrophysics Data System (ADS)

    Bodin, Christophe

    The calculation of the neutron flux is an important data that is used to determine the dynamic of the core of a Pressurized Water Reactor (PWR). However the transport equation which gives the neutron flux, cannot be solved in three dimensions over the whole core, in evolution because of the power of the current computers, which are too slow. So some simplifications are necessary to calculate this flux. Two-levels schemes are used, where, in a first step, some macroscopic cross sections libraries are generated by solving the transport equation using infinite lattice calculations on two dimensions assemblies. These sections are generally homogenized on the whole assembly and condensed to two energy groups. In a second step, the whole core calculation is carried out using the diffusion equation, with the cross sections of the libraries previously generated, interpolated at the values of the different parameters. However the core of a PWR is made up of many assemblies, that can contain two types of fuel : Uranium OXyde (UOX) or plutonium and uranium Mixed OXyde (MOX). Moreover all these assemblies have different burnup because each one can be used for three or four cycles depending on the PWR. So that imply some burnup gradients. Thus the hypothesis of the infinite lattice used to generate the cross sections libraries can be highly inaccurate. The first goal of this project is to generate cross sections libraries that take into account the environment and to evaluate the impact of this heterogeneous environment on the core calculation. The flux obtained with the diffusion equation at the end of the core calculation is not accurate enough, du to the homogenization by assembly, to determine and to locate the hotspot factor, which represents an important industrial problematic. The principle of the power reconstruction method (PRM) is to reconstruct the more accurately possible the flux in the pins, with a combination of the diffusion flux and some microscopic flux which

  4. Validation of updated neutronic calculation models proposed for Atucha-II PHWR. Part II: Benchmark comparisons of PUMA core parameters with MCNP5 and improvements due to a simple cell heterogeneity correction

    SciTech Connect

    Grant, C.; Mollerach, R.; Leszczynski, F.; Serra, O.; Marconi, J.; Fink, J.

    2006-07-01

    In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure vessel design with 451 vertical coolant channels and the fuel assemblies (FA) are clusters of 37 natural UO{sub 2} rods with an active length of 530 cm. For the reactor physics area, a revision and update of reactor physics calculation methods and models was recently carried out covering cell, supercell (control rod) and core calculations. This paper presents benchmark comparisons of core parameters of a slightly idealized model of the Atucha-I core obtained with the PUMA reactor code with MCNP5. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, more symmetric than Atucha-II, and has some experimental data available. To validate the new models benchmark comparisons of k-effective, channel power and axial power distributions obtained with PUMA and MCNP5 have been performed. In addition, a simple cell heterogeneity correction recently introduced in PUMA is presented, which improves significantly the agreement of calculated channel powers with MCNP5. To complete the validation, the calculation of some of the critical configurations of the Atucha-I reactor measured during the experiments performed at first criticality is also presented. (authors)

  5. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    SciTech Connect

    Wang, Jy-An John; Wang, Hong

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  6. Dry Storage Demonstration for High-Burnup Spent Nuclear Fuel-Feasibility Study

    SciTech Connect

    McKinnon, Mikal A. ); Cunningham, Mitchel E. )

    2003-09-09

    Initially, casks for dry storage of spent fuel were licensed for assembly-average burnup of about 35 GWd/MTU. Over the last two decades, the discharge burnup of fuel has increased steadily and now exceeds 45 GWd/MTU. With spent fuel burnups approaching the licensing limits (peak rod burnup of 62 GWd/MTU for pressurized water reactor fuel) and some lead test assemblies being burned beyond this limit, a need for a confirmatory dry storage demonstration program was first identified after the publication in May 1999 of the U.S. Nuclear Regulatory Commissions (NRC) Interim Staff Guidance 11 (ISG-11). With the publication in July 2002 of the second revision of ISG-11, the desirability for such a program further increased to obtain confirmatory data about the potential changes in cladding mechanical properties induced by dry storage, which would have implications to the transportation, handling, and disposal of high-burnup spent fuel. While dry storage licenses have kept pace with reactor discharge burnups, transportation licenses have not and are considered on a case by case basis. Therefore, this feasibility study was performed to examine the options available for conducting a confirmatory experimental program supporting the dry storage, transportation, and disposal of spent nuclear fuel with burnups well in excess of 45 GWd/MTU.

  7. Assessment of the use of extended burnup fuel in light water power reactors

    SciTech Connect

    Baker, D.A.; Bailey, W.J.; Beyer, C.E.; Bold, F.C.; Tawil, J.J.

    1988-02-01

    This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd/t uranium be increased to above 50 GWd/t. The environmental effects of extending fuel burnup during normal operations and during accident events and the economic effects of cost changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic assessments. Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction: (1) total fuel requirements; (2) reactor downtime for fuel replacement; (3) the number of fuel shipments to and from reactor sites; and (4) repository storage requirements. 61 refs., 4 figs., 27 tabs.

  8. Assessment of reactivity transient experiments with high burnup fuel

    SciTech Connect

    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.

  9. Temperature and burnup correlated fuel-cladding chemical interaction in U-10ZR metallic fuel

    NASA Astrophysics Data System (ADS)

    Carmack, William J.

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors and provide a number of advantages over other fuel types considering their fabricability, performance, recyclability, and safety. Resistance to cladding "breach" and subsequent release of fission products and fuel constituents to the nuclear power plant primary coolant system is a key performance parameter for a nuclear fuel system. In metallic fuel, FCCI weakens the cladding, especially at high power-high temperature operation, contributing to fuel pin breach. Empirical relationships for FCCI have been developed from a large body of data collected from in-pile (EBR-II) and out-of-pile experiments [1]. However, these relationships are unreliable in predicting FCCI outside the range of EBR-II experimental data. This dissertation examines new FCCI data extracted from the MFF-series of prototypic length metallic fuel irradiations performed in the Fast Flux Test Facility (FFTF). The fuel in these assemblies operated a temperature and burnup conditions similar to that in EBR-II but with axial fuel height three times longer than EBR-II experiments. Comparing FCCI formation data from FFTF and EBR-II provides new insight into FCCI formation kinetics. A model is developed combining both production and diffusion of lanthanides to the fuel-cladding interface and subsequent reaction with the cladding. The model allows these phenomena to be influenced by fuel burnup (lanthanide concentrations) and operating temperature. Parameters in the model are adjusted to reproduce measured FCCI layer thicknesses from EBR-II and FFTF. The model predicts that, under appropriate conditions, rate of FCCI formation can be controlled by either fission product transport or by the reaction rate of the interaction species at the fuel-cladding interface. This dissertation will help forward the design of metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full

  10. Features of the application of the Monte Carlo method to calculations for large RBMK reactors and to model correction on the basis of data from in-core detectors

    SciTech Connect

    Ivanov, I. E. Schukin, N. V.; Bychkov, S. A.; Druzhinin, V. E.; Lysov, D. A.; Shmonin, Yu. V.; Gurevich, M. I.

    2014-12-15

    Statistical errors in sampling neutron fields in physically large systems like an RBMK are analyzed both qualitatively and quantitatively. Recommendations concerning the choice of parameters for calculations are given. A new procedure for Monte Carlo RBMK calculations with model corrections on the basis of data from in-core detectors is proposed. Dedicated software based on the CUDA software and hardware platform is developed for computational research. Results of testing the procedure and software in question via calculations for real RBMK reactors are discussed.

  11. Analytical core loss calculations for magnetic materials used in high frequency high power converter applications. Ph.D. Thesis - Toledo Univ.

    NASA Technical Reports Server (NTRS)

    Triner, J. E.

    1979-01-01

    The basic magnetic properties under various operating conditions encountered in the state-of-the-art DC-AC/DC converters are examined. Using a novel core excitation circuit, the basic B-H and loss characteristics of various core materials may be observed as a function of circuit configuration, frequency of operation, input voltage, and pulse-width modulation conditions. From this empirical data, a mathematical loss characteristics equation is developed to analytically predict the specific core loss of several magnetic materials under various waveform excitation conditions.

  12. A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

    SciTech Connect

    Yang, W.S.; Kim, T.K.; Grandy, C.

    2007-07-01

    This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% {delta}k. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% {delta}k. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

  13. Monte Carlo simulation of the neutron characteristics of VVER-1000 core using the MCU-PD program and comparison of the results with calculations by the BIPR-7A program and experimental data

    SciTech Connect

    Dement'ev, V. G.; Oleinik, D. S.

    2011-12-15

    The Monte Carlo method has been used to simulate the neutron transport in nuclear reactors for over fifty years. Fast progress in computer power and development of more and more robust and reliable algorithms, codes, and nuclear databases allow solving more challenging problems, including three-dimensional (3D) simulations of full-scale reactor cores. Short descriptions of a full-scale 3D model of the VVER-1000 core and algorithms and methods implemented in the MCU-PD and BIPR-7A codes and a comparison of the calculations by each program as well as a comparison with experimental data are given in this paper.

  14. Development of Technical Basis for Burnup Credit Regulatory Guidance in the United States

    SciTech Connect

    Parks, Cecil V; Wagner, John C; Mueller, Don; Gauld, Ian C

    2011-01-01

    In the United States (U.S.) there has been and continues to be considerable interest in the increased use of burnup credit as part of the safety basis for SNF systems and this interest has motivated numerous technical studies related to the application of burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission initiated a burnup credit research program, with support from the Oak Ridge National Laboratory, to develop regulatory guidance and the supporting technical basis for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. The objective of this paper is to summarize the work and significant accomplishments, with references to the technical reports and publications for complete details.

  15. High burn-up structure of U(Mo) dispersion fuel

    NASA Astrophysics Data System (ADS)

    Leenaers, A.; Van Renterghem, W.; Van den Berghe, S.

    2016-08-01

    The evolution of the high burn-up structure (HBS) in U(Mo) fuel irradiated up to a burn-up of ∼70% 235U or ∼5 × 1021 f/cm3 or ∼120 GWd/tHM is described and compared to the observation made on LWR fuel. Scanning and transmission electron microscopy was performed on several samples having different burn-ups in order to get a better understanding of the mechanisms leading to the high burn-up structure formation. Even though there are some substantial differences between the irradiation of ceramic and U(Mo) alloy fuels (crystal structure, enrichment, irradiation temperature …), it was found that in both fuels recrystallization initiates at the same threshold and progresses in a similar way with increasing fission density. In case of U(Mo), recrystallization leads to accelerated swelling of the fuel which could result in instability of the fuel plate.

  16. Burnup of fusion produced tritons and /sup 3/He ions in PLT and PDX

    SciTech Connect

    Heidbrink, W.W.; Chrien, R.E.; Strachan, J.D.

    1982-09-01

    The d(d,p)t and d(d,n)/sup 3/He fusion reactions produce 1 MeV tritons and 0.8 MeV /sup 3/He ions which can subsequently undergo d(t,n)..cap alpha.. and d(/sup 3/He,p)..cap alpha.. fusion reactions. The magnitude of this triton and /sup 3/He ion burnup was measured on the PLT and PDX tokamaks by detection of the 14 MeV neutron and 15 MeV proton emission. In discharges with B/sub phi/ greater than or equal to 2 T, the measured /sup 3/He burnup agrees well with predictions based on classical theories of ion confinement and slowing down, while the triton burnup was about four times lower than theoretically predicted. In discharges with weaker toroidal fields, the burnup of both ions fell by more than a factor of ten.

  17. French investigations of high burnup effect on LOCA thermomecanical behavior. Part two. Oxidation and quenching experiments under simulated LOCA conditions with high burnup clad material

    SciTech Connect

    GrandJean, C.; Cauvin, R.; Lebuffe, C.

    1997-01-01

    In the frame of the high burnup fuel studies to support a possible extension of the current discharge burnup limit, experimental programs have been undertaken, jointly by EDF and IPSN in order to study the thermal-shock behavior of high burnup fuel claddings under typical LOCA conditions. The TAGUS program used unirradiated cladding samples, bare or bearing a pre-corrosion state simulating the end-of-life state of high burnup fuel claddings: the TAGCIR program used actually irradiated cladding samples taken from high burnup rods irradiated over 5 cycles in a commercial EDF PWR and having reached a rod burnup close to 60 GWd/tU. The thermal-shock failure tests consisted in oxidizing the cladding samples under steam flow, on both inner and outer faces or on the outer face alone, and subjecting them to a final water quench. The heating was provided by an inductive furnace the power of which being regulated through monitoring of the sample surface temperature with use of a single-wave optical pyrometer. Analysis of the irradiated tests (TAGCIR series) evidenced an increased oxidation rate as compared to similar tests on unirradiated samples. Results of the quenching tests series on unirradiated and irradiated samples are plotted under the usual presentation of failure maps relative to the oxidation parameters ECR (equivalent cladding reacted) or e{sub {beta}} (thickness of the remaining beta phase layer) as a function of the oxidation temperature. Comparison of the failure limits for irradiated specimens to those for unirradiated specimens indicates a lower brittleness under two side oxidation and possibly the opposite under one-side oxidation. The tentative analysis of the oxidation and quenching tests results on irradiated samples reveals the important role played by the hydrogen charged during in-reactor corrosion on the oxidation kinetics and the failure bearing capability of the cladding under LOCA transient conditions.

  18. High Burnup Effects Program A State-of-the-Technology Assessment

    SciTech Connect

    Rising, K. H.; Bradley, E. R.; Williford, R. E.; Freshley, M D.

    1982-06-01

    Various analytical models and empirical correlations describing the fission gas release phenomenon were examined. An evaluation was made of the current pertinent experimental data on the subject of high burnup fission gas release. Data reported by individual investigators were compared and evaluated in relation to their applicability to the content and scope of the High Burnup Effects Program. These evaluations then form the bases for defining the data needs, and the selection of variables to be studied in this program.

  19. LWR fuel-cycle costs as a function of burnup. Final report

    SciTech Connect

    Franks, W.; Goldstein, L.; Joseph, L.; Nikmohammadian, N.

    1984-11-01

    Utilities may be able to decrease fuel-cycle costs as much as 5% in PWRs and 6% in BWRs by increasing discharge burnup to optimum practical limits. With one exception, this analysis of 12- and 18-month fuel cycles indicated a potential for still further cost reductions at higher burnup rates than those considered (39 GWd/MtU for BWRs and 55 GWd/MtU for PWRs).

  20. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses--Criticality (keff) Predictions

    SciTech Connect

    Scaglione, John M; Mueller, Don; Wagner, John C

    2011-01-01

    One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission

  1. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    NASA Astrophysics Data System (ADS)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  2. Cladding stress during extended storage of high burnup spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Raynaud, Patrick A. C.; Einziger, Robert E.

    2015-09-01

    In an effort to assess the potential for low temperature creep and delayed hydride cracking failures in high burnup spent fuel cladding during extended dry storage, the U.S. NRC analytical fuel performance tools were used to predict cladding stress during a 300 year dry storage period for UO2 fuel burned up to 65 GWd/MTU. Fuel swelling correlations were developed and used along with decay gas production and release fractions to produce circumferential average cladding stress predictions with the FRAPCON-3.5 fuel performance code. The resulting stresses did not result in cladding creep failures. The maximum creep strains accumulated were on the order of 0.54-1.04%, but creep failures are not expected below at least 2% strain. The potential for delayed hydride cracking was assessed by calculating the critical flaw size required to trigger this failure mechanism. The critical flaw size far exceeded any realistic flaw expected in spent fuel at end of reactor life.

  3. A Two-Step Approach to Uncertainty Quantification of Core Simulators

    SciTech Connect

    Yankov, Artem; Collins, Benjamin; Klein, Markus; Jessee, Matthew A.; Zwermann, Winfried; Velkov, Kiril; Pautz, Andreas; Downar, Thomas

    2012-01-01

    For the multiple sources of error introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily to quantify the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the “two-step” method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the two-step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor and in the core power distribution was examined in the framework of phase I-3 of the OECD Uncertainty Analysis in Modeling benchmark. With the Three Mile Island Unit 1 core as a base model for analysis, the XSUSA and two-step methods were applied with certain limitations, and the results were compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions were drawn as to the method that is currently more viable for computing uncertainties in burnup and transient calculations.

  4. A Two-Step Approach to Uncertainty Quantification of Core Simulators

    DOE PAGES

    Yankov, Artem; Collins, Benjamin; Klein, Markus; Jessee, Matthew A.; Zwermann, Winfried; Velkov, Kiril; Pautz, Andreas; Downar, Thomas

    2012-01-01

    For the multiple sources of error introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily to quantify the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the “two-step” method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the two-step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor andmore » in the core power distribution was examined in the framework of phase I-3 of the OECD Uncertainty Analysis in Modeling benchmark. With the Three Mile Island Unit 1 core as a base model for analysis, the XSUSA and two-step methods were applied with certain limitations, and the results were compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions were drawn as to the method that is currently more viable for computing uncertainties in burnup and transient calculations.« less

  5. VORCOR: A computer program for calculating characteristics of wings with edge vortex separation by using a vortex-filament and-core model

    NASA Technical Reports Server (NTRS)

    Pao, J. L.; Mehrotra, S. C.; Lan, C. E.

    1982-01-01

    A computer code base on an improved vortex filament/vortex core method for predicting aerodynamic characteristics of slender wings with edge vortex separations is developed. The code is applicable to camber wings, straked wings or wings with leading edge vortex flaps at subsonic speeds. The prediction of lifting pressure distribution and the computer time are improved by using a pair of concentrated vortex cores above the wing surface. The main features of this computer program are: (1) arbitrary camber shape may be defined and an option for exactly defining leading edge flap geometry is also provided; (2) the side edge vortex system is incorporated.

  6. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Jiang, Hao; Bevard, Bruce Balkcom; Howard, Rob L; Scaglione, John M

    2015-01-01

    The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into local stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

  7. Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

    SciTech Connect

    Enercon Services, Inc.

    2011-03-14

    Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in

  8. Computational model for calculating body-core temperature elevation in rabbits due to whole-body exposure at 2.45 GHz

    NASA Astrophysics Data System (ADS)

    Hirata, Akimasa; Sugiyama, Hironori; Kojima, Masami; Kawai, Hiroki; Yamashiro, Yoko; Fujiwara, Osamu; Watanabe, Soichi; Sasaki, Kazuyuki

    2008-06-01

    In the current international guidelines and standards with regard to human exposure to electromagnetic waves, the basic restriction is defined in terms of the whole-body average-specific absorption rate. The rationale for the guidelines is that the characteristic pattern of thermoregulatory response is observed for the whole-body average SAR above a certain level. However, the relationship between energy absorption and temperature elevation was not well quantified. In this study, we improved our thermal computation model for rabbits, which was developed for localized exposure on eye, in order to investigate the body-core temperature elevation due to whole-body exposure at 2.45 GHz. The effect of anesthesia on the body-core temperature elevation was also discussed in comparison with measured results. For the whole-body average SAR of 3.0 W kg-1, the body-core temperature in rabbits elevates with time, without becoming saturated. The administration of anesthesia suppressed body-core temperature elevation, which is attributed to the reduced basal metabolic rate.

  9. Core Optimization of a Deep-Burn Pebble Bed Reactor

    SciTech Connect

    Brian Boer; Abderrafi M. Ougouag

    2010-06-01

    Achieving a high fuel burnup in the Deep-Burn (DB) pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum as compared to a ’standard’ UO2 fueled core. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. The DB concept focuses on the destruction of spent fuel transuranics in TRISO coated particle fueled gas-cooled reactors with the aim of a fractional fuel burnup of 60-70% in fissions per initial metal atom (FIMA), using a single-pass, multi in-core fuel (re)cycling scheme. In principle, the DB pebble bed concept employs the same reactor designs as the present low enriched uranium core designs, i.e. the 400 MWth Pebble Bed Modular Reactor (PBMR-400). A Pu and Minor Actinide fueled PBMR-400 design serves as the starting point for a core optimization study. The fuel temperature, power peak, temperature reactivity coefficients, and burnup capabilities of the modified designs are analyzed with the PEBBED code. A code-to-code coupling with the PASTA code allows for the analysis of the TRISO fuel performance for both normal and Loss Of Forced Cooling conditions. An improved core design is sought, maximizing the fuel discharge burnup, while retaining negative temperature reactivity feedback coefficients for the entire temperature range and avoiding high fuel temperatures (fuel failure probabilities).

  10. New high burnup fuel models for NRC`s licensing audit code, FRAPCON

    SciTech Connect

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1996-03-01

    Fuel behavior models have recently been updated within the U.S. Nuclear Regulatory Commission steady-state FRAPCON code used for auditing of fuel vendor/utility-codes and analyses. These modeling updates have concentrated on providing a best estimate prediction of steady-state fuel behavior up to the maximum burnup level s of current data (60 to 65 GWd/MTU rod-average). A decade has passed since these models were last updated. Currently, some U.S. utilities and fuel vendors are requesting approval for rod-average burnups greater than 60 GWd/MTU; however, until these recent updates the NRC did not have valid fuel performance models at these higher burnup levels. Pacific Northwest Laboratory (PNL) has reviewed 15 separate effects models within the FRAPCON fuel performance code (References 1 and 2) and identified nine models that needed updating for improved prediction of fuel behavior at high burnup levels. The six separate effects models not updated were the cladding thermal properties, cladding thermal expansion, cladding creepdown, fuel specific heat, fuel thermal expansion and open gap conductance. Comparison of these models to the currently available data indicates that these models still adequately predict the data within data uncertainties. The nine models identified as needing improvement for predicting high-burnup behavior are fission gas release (FGR), fuel thermal conductivity (accounting for both high burnup effects and burnable poison additions), fuel swelling, fuel relocation, radial power distribution, fuel-cladding contact gap conductance, cladding corrosion, cladding mechanical properties and cladding axial growth. Each of the updated models will be described in the following sections and the model predictions will be compared to currently available high burnup data.

  11. Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

    SciTech Connect

    Tiberi, V.

    2012-07-01

    The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity of the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine

  12. Investigation of the Effect of Fixed Absorbers on the Reactivity of PWR Spent Nuclear Fuel for Burnup Credit

    SciTech Connect

    Wagner, John C.; Sanders, Charlotta E.

    2002-08-15

    The effect of fixed absorbers on the reactivity of pressurized water reactor (PWR) spent nuclear fuel (SNF) in support of burnup-credit criticality safety analyses is examined. A fuel assembly burned in conjunction with fixed absorbers may have a higher reactivity for a given burnup than an assembly that has not used fixed absorbers. As a result, guidance on burnup credit, issued by the U.S. Nuclear Regulatory Commission's Spent Fuel Project Office, recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommendation eliminates a large portion of the currently discharged SNF from loading in burnup credit casks and thus severely limits the practical usefulness of burnup credit. Therefore, data are needed to support the extension of burnup credit to additional SNF. This research investigates the effect of various fixed absorbers, including integral burnable absorbers, burnable poison rods, control rods, and axial power shaping rods, on the reactivity of PWR SNF. Trends in reactivity with relevant parameters (e.g., initial fuel enrichment, burnup and absorber type, exposure, and design) are established, and anticipated reactivity effects are quantified. Where appropriate, recommendations are offered for addressing the reactivity effects of the fixed absorbers in burnup-credit safety analyses.

  13. Core-core and core-valence correlation

    NASA Technical Reports Server (NTRS)

    Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

    1988-01-01

    The effect of (1s) core correlation on properties and energy separations was analyzed using full configuration-interaction (FCI) calculations. The Be 1 S - 1 P, the C 3 P - 5 S and CH+ 1 Sigma + or - 1 Pi separations, and CH+ spectroscopic constants, dipole moment and 1 Sigma + - 1 Pi transition dipole moment were studied. The results of the FCI calculations are compared to those obtained using approximate methods. In addition, the generation of atomic natural orbital (ANO) basis sets, as a method for contracting a primitive basis set for both valence and core correlation, is discussed. When both core-core and core-valence correlation are included in the calculation, no suitable truncated CI approach consistently reproduces the FCI, and contraction of the basis set is very difficult. If the (nearly constant) core-core correlation is eliminated, and only the core-valence correlation is included, CASSCF/MRCI approached reproduce the FCI results and basis set contraction is significantly easier.

  14. Radionuclide Data and Calculations and Loss-On-Ignition, X-Ray Fluorescence, and ICP-AES Data from Cores in Catchments of the Animas River, Colorado

    USGS Publications Warehouse

    Church, Stanley E.; Rice, Cyndi A.; Marot, Marci E.

    2008-01-01

    The U.S. Departments of Agriculture and Interior Abandoned Mine Lands (AML) Initiative is focused on the evaluation of the effect of past mining practices on the water quality and the riparian and aquatic habitats of impacted stream reaches downstream from historical mining districts located primarily on Federal lands. This problem is manifest in the eleven western states (west of longitude 102 degrees) where the majority of hardrock mines that had past production are located on Federal lands. In areas of temperate climate and moderate to heavy precipitation, the effects of rapid chemical and physical weathering of sulfides exposed on mine-waste dumps and acidic drainage from mines have resulted in elevated metal concentrations in the stream water and stream-bed sediment. The result of these mineral weathering processes has an unquantified impact on the quality of the water and the aquatic and riparian habitats that may limit their recreational resource value. One of the confounding factors in these studies is the determination of the component of metals derived from hydrothermally altered but unmined portions of these drainage basins. Several watersheds have been studied to evaluate the effects of acid mine drainage and acid rock drainage on the near-surface environment. The Animas River watershed in southwestern Colorado contains a large number of past-producing metal mines that have affected the watershed. Beginning in October 1996, the U.S. Geological Survey (USGS) began a collaborative study of these effects under the USGS-AML Initiative. In this report, we present the radionuclide and geochemical analytical results of sediment coring during 1997-1999 from two cores from oxbow lakes 0.5 mi. upstream from the 32nd Street Bridge near Durango, Colo., and from three cores from beaver ponds within the Mineral Creek drainage basin near Silverton, Colo.

  15. Hybrid Reduced Order Modeling Algorithms for Reactor Physics Calculations

    NASA Astrophysics Data System (ADS)

    Bang, Youngsuk

    Reduced order modeling (ROM) has been recognized as an indispensable approach when the engineering analysis requires many executions of high fidelity simulation codes. Examples of such engineering analyses in nuclear reactor core calculations, representing the focus of this dissertation, include the functionalization of the homogenized few-group cross-sections in terms of the various core conditions, e.g. burn-up, fuel enrichment, temperature, etc. This is done via assembly calculations which are executed many times to generate the required functionalization for use in the downstream core calculations. Other examples are sensitivity analysis used to determine important core attribute variations due to input parameter variations, and uncertainty quantification employed to estimate core attribute uncertainties originating from input parameter uncertainties. ROM constructs a surrogate model with quantifiable accuracy which can replace the original code for subsequent engineering analysis calculations. This is achieved by reducing the effective dimensionality of the input parameter, the state variable, or the output response spaces, by projection onto the so-called active subspaces. Confining the variations to the active subspace allows one to construct an ROM model of reduced complexity which can be solved more efficiently. This dissertation introduces a new algorithm to render reduction with the reduction errors bounded based on a user-defined error tolerance which represents the main challenge of existing ROM techniques. Bounding the error is the key to ensuring that the constructed ROM models are robust for all possible applications. Providing such error bounds represents one of the algorithmic contributions of this dissertation to the ROM state-of-the-art. Recognizing that ROM techniques have been developed to render reduction at different levels, e.g. the input parameter space, the state space, and the response space, this dissertation offers a set of novel

  16. Monte Carlo and theoretical calculations of the first four perturbation coefficients in the high temperature series expansion of the free energy for discrete and core-softened potential models

    NASA Astrophysics Data System (ADS)

    Zhou, Shiqi; Solana, J. R.

    2013-06-01

    The first four perturbation coefficients in the expansion of the Helmholtz free energy in power series of the inverse of the reduced temperature for a number of potential models with hard-sphere cores plus core-softened and discontinuous tails are obtained from Monte Carlo simulations. The potential models considered include square-well, double square-well, and square-shoulder plus square-well, with different potential parameters. These simulation data are used to evaluate the performance of a traditional macroscopic compressibility approximation (MCA) for the second order coefficient and a recent coupling parameter series expansion (CPSE) for the first four coefficients. Comprehensive comparison indicates the incapability of the MCA for the second order coefficient in most non-stringent situations, and significance of the CPSE in accurately calculating these four coefficients.

  17. Monte Carlo and theoretical calculations of the first four perturbation coefficients in the high temperature series expansion of the free energy for discrete and core-softened potential models.

    PubMed

    Zhou, Shiqi; Solana, J R

    2013-06-28

    The first four perturbation coefficients in the expansion of the Helmholtz free energy in power series of the inverse of the reduced temperature for a number of potential models with hard-sphere cores plus core-softened and discontinuous tails are obtained from Monte Carlo simulations. The potential models considered include square-well, double square-well, and square-shoulder plus square-well, with different potential parameters. These simulation data are used to evaluate the performance of a traditional macroscopic compressibility approximation (MCA) for the second order coefficient and a recent coupling parameter series expansion (CPSE) for the first four coefficients. Comprehensive comparison indicates the incapability of the MCA for the second order coefficient in most non-stringent situations, and significance of the CPSE in accurately calculating these four coefficients.

  18. Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

    SciTech Connect

    William J. Carmack

    2012-05-01

    the EBR-II and study of the differences between the two fuel systems is critical for design of large advanced sodium cooled fast reactor systems. Comparing FCCI layer formation data between FFTF and EBR-II indicates that the same diffusion model can be used to represent the two systems when considering time, temperature, burnup history, and axial temperature and power profiles. This dissertation shows that FCCI formation peaks further below the top of the fuel column in FFTF experiments than has been observed in EBR-II experiments. The work provided in this dissertation will help forward the design of advanced metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full length reactor designs. This will allow the accurate lifetime prediction of fuel performance capability for new advanced sodium cooled fast reactors with extended core designs.

  19. Determination of curie content and {sup 134/137}cesium ratios by gamma spectroscopy of high burnup plutonium-aluminum fuel assemblies

    SciTech Connect

    Haggard, D.L.; Tanner, J.E.

    1997-06-01

    Nondestructive assay (NDA) gamma spectroscopy techniques were used to measure {sup 134/137}Cs ratios on nine PuAl Mark 42 fuel assemblies. The purpose of the ratio measurement was to confirm theoretical burnup calculations. {sup 134/137}Cs ratios were determined from the measured activity based on corrected net peak area counts for the 605 keV peak from {sup 134}Cs and the 662 keV peak from {sup 137}Cs/{sup 137m}Ba. Assembly No. 2 {sup 134/137}Cs ratio measured on 4-15-92 was 0.19. The measured {sup 134/137}Cs ratio was decay corrected to be 2.11 on 8-1-84 based on the half lives of {sup 134}Cs and {sup 137}Cs. The measured {sup 134/137}Cs ratio range was 1.90--2.14 for all nine assemblies. These measured values compare to a theoretical ratio of 1.7 on 8-1-84 determined by burnup calculations. Total cesium curie content was also requested and determined using the NDA direct measurements. Gamma spectral data were measured on the nine sectioned Mark 42 fuel assemblies. Measured cesium curie content, decay corrected to 8-1-84, ranged from 18170--24480 curies of {sup 134}Cs and 8620--11646 curies of {sup 137}Cs. Theoretical cesium curie content of 8-1-84 was 15200 curies {sup 134}Cs and 8973 curies {sup 137}Cs. Direct assay cesium ratio is 12% to 26% higher than the predicted ratio of 1.7. The measured {sup 134}Cs data indicate between 20%--61% more activity than that predicted by the burnup code, whereas the measured {sup 137}Cs activity is between 4% less to 30% more than the predicted activity. This information may be used to address issues concerning criticality safety, storage, and shipping of this type of material.

  20. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses: Criticality (keff) Predictions

    DOE PAGES

    Scaglione, John M.; Mueller, Don E.; Wagner, John C.

    2014-12-01

    One of the most important remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. In this study, this paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (keff) evaluations based on best-available data and methodsmore » and applies the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The criticality validation approach utilizes not only available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion program to support validation of the principal actinides but also calculated sensitivities, nuclear data uncertainties, and limited available FP LCE data to predict and verify individual biases for relevant minor actinides and FPs. The results demonstrate that (a) sufficient critical experiment data exist to adequately validate keff calculations via conventional validation approaches for the primary actinides, (b) sensitivity-based critical experiment selection is more appropriate for generating accurate application model bias and uncertainty, and (c) calculated sensitivities and nuclear data uncertainties can be used for generating conservative estimates of bias for minor actinides and FPs. Results based on the SCALE 6.1 and the ENDF/B-VII.0 cross-section libraries indicate that a conservative estimate of the bias for the minor actinides and FPs is 1.5% of their worth within the

  1. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    SciTech Connect

    Wiesenack, W.

    1996-03-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project`s data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup.

  2. 78 FR 67348 - Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-12

    ... the structures, systems, and components of the dry cask storage systems for high burnup spent nuclear... of dry storage cask systems and experiments would be beneficial. A draft test plan for the High... Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

  3. International studies on burnup credit criticality safety by an OECD/NEA working group

    SciTech Connect

    Brady, M.C.; Okuno, H.; DeHart, M.D.; Nouri, A.; Sartori, E.

    1998-11-01

    The results and conclusions from a six-year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses is presented. Approximately 20 participants from 12 countries have provided results for most problems. Four detailed benchmark problems for pressurized-water-reactor fuel have been completed. Results from work being finalized, addressing burnup credit for boiling-water-reactor fuel, are discussed, as well as planned activities for additional benchmarks, including mixed-oxide fuels, and other activities.

  4. Progress of the RIA experiments with high burnup fuels and their evaluation in JAERI

    SciTech Connect

    Ishijima, Kiyomi; Fuketa, Toyoshi

    1997-01-01

    Recent results obtained in the NSRR power burst experiments with high burnup PWR fuel rods are described and discussed in this paper. Data concerning test condition, transient records during pulse irradiation and post irradiation examination are described. Another high burnup PWR fuel rod failed in the test HBO-5 at the slightly higher energy deposition than that in the test HBO-1. The failure mechanism of the test HBO-5 is the same as that of the test HBO-1, that is, hydride-assisted PCMI. Some influence of the thermocouples welding on the failure behavior of the HBO-5 rod was observed.

  5. Flux reconstruction methods for assembly calculations in the code APOLLO2

    SciTech Connect

    Zmijarevic, I.; Masiello, E.; Sanchez, R.

    2006-07-01

    A technique for flux reconstruction has been incorporated in the code APOLLO2 allowing for fast generation of accurate burnup libraries. The burnup flux is obtained as the product of a pivot flux from a large-macro-group heterogeneous-assembly calculation times an energy shape factor based on a fine-energy flux. The latter is obtained from a fast multicell calculation for a few types of representative cells in the assembly. The reconstruction formula preserves the reaction rates predicted by the large-macro-group, heterogeneous-assembly calculation. Analysis of a BWR MOX benchmark shows that the reactivity error for a 70 MWd/t burnup cycle did not exceed 90 pcm and the maximum error in pin powers did not exceed 1%. (authors)

  6. Results of a Neutronic Simulation of HTR-Proteus Core 4.2 using PEBBED and other INL Reactor Physics Tools: FY-09 Report

    SciTech Connect

    Hans D. Gougar

    2009-08-01

    The Idaho National Laboratory’s deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. A combination of unit cell calculations (COMBINE-PEBDAN), 1-D discrete ordinates transport (SCAMP), and nodal diffusion calculations (PEBBED) were employed to yield keff and flux profiles. Preliminary results indicate that these tools, as currently configured and used, do not yield satisfactory estimates of keff. If control rods are not modeled, these methods can deliver much better agreement with experimental core eigenvalues which suggests that development efforts should focus on modeling control rod and other absorber regions. Under some assumptions and in 1D subcore analyses, diffusion theory agrees well with transport. This suggests that developments in specific areas can produce a viable core simulation approach. Some corrections have been identified and can be further developed, specifically: treatment of the upper void region, treatment of inter-pebble streaming, and explicit (multiscale) transport modeling of TRISO fuel particles as a first step in cross section generation. Until corrections are made that yield better agreement with experiment, conclusions from core design and burnup analyses should be regarded as qualitative and not benchmark quality.

  7. Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

    SciTech Connect

    Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F.

    2013-07-01

    This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

  8. Estimation of average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors by using the {sup 134}Cs/{sup 137}Cs ratio method

    SciTech Connect

    Endo, T.; Sato, S.; Yamamoto, A.

    2012-07-01

    Average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors is estimated, using the {sup 134}Cs/{sup 137}Cs ratio method for measured radioactivities of {sup 134}Cs and {sup 137}Cs in contaminated soils within the range of 100 km from the Fukushima Dai-ichi nuclear power plants. As a result, the measured {sup 134}Cs/{sup 137}Cs ratio from the contaminated soil is 0.996{+-}0.07 as of March 11, 2011. Based on the {sup 134}Cs/{sup 137}Cs ratio method, the estimated burnup of damaged fuels is approximately 17.2{+-}1.5 [GWd/tHM]. It is noted that the numerical results of various calculation codes (SRAC2006/PIJ, SCALE6.0/TRITON, and MVP-BURN) are almost the same evaluation values of {sup 134}Cs/ {sup 137}Cs ratio with same evaluated nuclear data library (ENDF-B/VII.0). The void fraction effect in depletion calculation has a major impact on {sup 134}Cs/{sup 137}Cs ratio compared with the differences between JENDL-4.0 and ENDF-B/VII.0. (authors)

  9. Measurement and calculation of the fast-neutron and photon spectra from the core boundary to the biological shielding in the WWER-1000 reactor model.

    PubMed

    Osmera, B; Cvachovec, F; Kyncl, J; Smutný, V

    2005-01-01

    The fast-neutron and photon space-energy distributions have been measured in an axially (1.25 m active height) and azimuthally (60 degree symmetry sector) shortened model of the WWER-1000 reactor assembled in the LR-0 experimental reactor. The space-energy distributions have been calculated with the stochastic code MCNP and the deterministic three-dimensional code TORT. Selected results are presented and discussed in the paper. This work has been done in the frame of the EU 5th FW project REDOS REDOS, Reactor Dosimetry: Accurate determination and benchmarking of radiation field parameters, relevant for reactor pressure vessel monitoring. EURATOM Programme, Call 2000/C 294/04). All geometry and material composition data of the model as well as the available experimental data were carefully checked and revised.

  10. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels

    SciTech Connect

    Bai, Xian -Ming; Tonks, Michael R.; Zhang, Yongfeng; Hales, Jason D.

    2015-12-22

    The high burnup structure forming at the rim region in UO2 based nuclear fuel pellets has interesting physical properties such as improved thermal conductivity, even though it contains a high density of grain boundaries and micron-size gas bubbles. To understand this counterintuitive phenomenon, mesoscale heat conduction simulations with inputs from atomistic simulations and experiments were conducted to study the thermal conductivities of a small-grain high burnup microstructure and two large-grain unrestructured microstructures. We concluded that the phonon scattering effects caused by small point defects such as dispersed Xe atoms in the grain interior must be included in order to correctly predict the thermal transport properties of these microstructures. In extreme cases, even a small concentration of dispersed Xe atoms such as 10-5 can result in a lower thermal conductivity in the large-grain unrestructured microstructures than in the small-grain high burnup structure. The high-density grain boundaries in a high burnup structure act as defect sinks and can reduce the concentration of point defects in its grain interior and improve its thermal conductivity in comparison with its large-grain counterparts. Furthermore, an analytical model was developed to describe the thermal conductivity at different concentrations of dispersed Xe, bubble porosities, and grain sizes. Upon calibration, the model is robust and agrees well with independent heat conduction modeling over a wide range of microstructural parameters.

  11. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels

    DOE PAGES

    Bai, Xian -Ming; Tonks, Michael R.; Zhang, Yongfeng; Hales, Jason D.

    2015-12-22

    The high burnup structure forming at the rim region in UO2 based nuclear fuel pellets has interesting physical properties such as improved thermal conductivity, even though it contains a high density of grain boundaries and micron-size gas bubbles. To understand this counterintuitive phenomenon, mesoscale heat conduction simulations with inputs from atomistic simulations and experiments were conducted to study the thermal conductivities of a small-grain high burnup microstructure and two large-grain unrestructured microstructures. We concluded that the phonon scattering effects caused by small point defects such as dispersed Xe atoms in the grain interior must be included in order to correctlymore » predict the thermal transport properties of these microstructures. In extreme cases, even a small concentration of dispersed Xe atoms such as 10-5 can result in a lower thermal conductivity in the large-grain unrestructured microstructures than in the small-grain high burnup structure. The high-density grain boundaries in a high burnup structure act as defect sinks and can reduce the concentration of point defects in its grain interior and improve its thermal conductivity in comparison with its large-grain counterparts. Furthermore, an analytical model was developed to describe the thermal conductivity at different concentrations of dispersed Xe, bubble porosities, and grain sizes. Upon calibration, the model is robust and agrees well with independent heat conduction modeling over a wide range of microstructural parameters.« less

  12. Multiscale modeling of thermal conductivity of high burnup structures in UO2 fuels

    NASA Astrophysics Data System (ADS)

    Bai, Xian-Ming; Tonks, Michael R.; Zhang, Yongfeng; Hales, Jason D.

    2016-03-01

    The high burnup structure forming at the rim region in UO2 based nuclear fuel pellets has interesting physical properties such as improved thermal conductivity, even though it contains a high density of grain boundaries and micron-size gas bubbles. To understand this counterintuitive phenomenon, mesoscale heat conduction simulations with inputs from atomistic simulations and experiments were conducted to study the thermal conductivities of a small-grain high burnup microstructure and two large-grain unrestructured microstructures. We concluded that the phonon scattering effects caused by small point defects such as dispersed Xe atoms in the grain interior must be included in order to correctly predict the thermal transport properties of these microstructures. In extreme cases, even a small concentration of dispersed Xe atoms such as 10-5 can result in a lower thermal conductivity in the large-grain unrestructured microstructures than in the small-grain high burnup structure. The high-density grain boundaries in a high burnup structure act as defect sinks and can reduce the concentration of point defects in its grain interior and improve its thermal conductivity in comparison with its large-grain counterparts. An analytical model was developed to describe the thermal conductivity at different concentrations of dispersed Xe, bubble porosities, and grain sizes. Upon calibration, the model is robust and agrees well with independent heat conduction modeling over a wide range of microstructural parameters.

  13. Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

    SciTech Connect

    Gauld, I.C.

    2005-08-12

    U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k{sub eff}) to determine the net importance of cross sections to k{sub eff}. The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: {sup 151}Sm, {sup 103}Rh, {sup 155}Eu, {sup 150}Sm, {sup 152}Sm, {sup 153}Eu, {sup 154}Eu, and {sup 143}Nd.

  14. Microhardness and Young's modulus of high burn-up UO2 fuel

    NASA Astrophysics Data System (ADS)

    Cappia, F.; Pizzocri, D.; Marchetti, M.; Schubert, A.; Van Uffelen, P.; Luzzi, L.; Papaioannou, D.; Macián-Juan, R.; Rondinella, V. V.

    2016-10-01

    Vickers microhardness (HV0.1) and Young's modulus (E) measurements of LWR UO2 fuel at burn-up ≥60 GWd/tHM are presented. Their ratio HV0.1/E was found constant in the range 60-110 GWd/tHM. From the ratio and the microhardness values vs porosity, the Young's modulus dependence on porosity was derived and extended to the full radial profile, including the high burn-up structure (HBS). The dependence is well represented by a linear correlation. The data were compared to fuel performance codes correlations. A burn-up dependent factor was introduced in the Young's modulus expression. The modifications extend the experimental validation range of the TRANSURANUS correlation from un-irradiated to irradiated UO2 and up to 20% porosity. First simulations of LWR fuel rod irradiations were performed in order to illustrate the impact on fuel performance. In the specific cases selected, the simulations suggest a limited effect of the Young's modulus decrease due to burn-up on integral fuel performance.

  15. TRU transmutation in thorium-based heterogeneous PWR core

    SciTech Connect

    Bae, Kang-Mok; Lim, Jae-Yong; Kim, Myung-Hyun

    2004-07-01

    A thorium-based seed and blanket design concept for a conventional pressurized light water reactor (PWR) was proposed to enhance the proliferation resistance potential and fuel cycle economics. The KTF core was satisfied with neutronic and thermal-hydraulic design limit of conventional PWR, APR-1400. In order to evaluate transmutation capability of a thorium-based KTF core, U/Zr seed fuel mixed with 10% TRU which come from 1,000 MWe power reactor after 10 years decay was proposed and analyzed by transmutation indices such as D{sub j}, TEX and SR. KTF core showed an extended fuel cycle burnup; average burnup of seed was 79.5 MWd/kgHM and blanket was 94.6 MWd/kgHM. It means that residence time of TRU in the core could be long enough for transmutation when TRU is mixed in seed fuel. The amount of TRU production from conventional PWR could be transmuted in the KTF-TRU core, especially Am-241 isotope is remarkably transmuted by capture reaction. Even isotopes of curium were cumulated in the core during the burnup, however, KTF-TRU core could reduce the TRU in spent fuel by using well-thermalized neutron spectrum. Proliferation resistance potential of thorium based transmutation fuel is slightly increased. About 31% reduction of TRU amount was measured from reduced plutonium production from U-238. Total amount of Am-241 was reduced significantly, but total amount of minor actinide (MA) was reduced by 28% of its initial loading mass. (authors)

  16. Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core

    SciTech Connect

    Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G.

    2006-12-15

    Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of {sub 62}{sup 149}Sm and its dependence on the shift of a resonance position E{sub r} (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73{<=}{delta}E{sub r}{<=}62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant {alpha}. We obtain new, more accurate limits of -4x10{sup -17}{<=}{alpha}{center_dot}/{alpha}{<=}3x10{sup -17} yr{sup -1}. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

  17. Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core

    NASA Astrophysics Data System (ADS)

    Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G.

    2006-12-01

    Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of 62149Sm and its dependence on the shift of a resonance position Er (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73⩽ΔEr⩽62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant α. We obtain new, more accurate limits of -4×10-17⩽α·/α⩽3×10-17yr-1. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

  18. Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

    SciTech Connect

    G. S. Chang

    2006-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

  19. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    SciTech Connect

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-07-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at {approx}2.4, {approx}7 and {approx}11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of {approx}7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10{sup 15} n/cm{sup 2}/s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between {approx}410 deg. C and {approx}645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  20. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    SciTech Connect

    2014-02-27

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

  1. Core-core and core-valence correlation

    NASA Technical Reports Server (NTRS)

    Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

    1988-01-01

    The effect of 1s core correlation on properties and energy separations are analyzed using full configuration-interaction (FCI) calculations. The Be1S - 1P, the C 3P - 5S,m and CH(+) 1Sigma(+) - 1Pi separations, and CH(+) spectroscopic constants, dipole moment, and 1Sigma(+) - 1Pi transition dipole moment have been studied. The results of the FCI calculations are compared to those obtained using approximate methods.

  2. Behaviour of fission gas in the rim region of high burn-up UO 2 fuel pellets with particular reference to results from an XRF investigation

    NASA Astrophysics Data System (ADS)

    Mogensen, M.; Pearce, J. H.; Walker, C. T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.

  3. AN Core Analysis

    NASA Astrophysics Data System (ADS)

    Barbarino, Andrea; Tomatis, Daniele

    2014-06-01

    Several alternative approximations of neutron transport have been proposed in years to move around the known limitations imposed by neutron diffusion in the modeling of nuclear cores. However, only a few complied with the industrial requirements of fast numerical computation, concentrating more on physical accuracy. In this work, the AN transport methodology is discussed with particular interest in core performance calculations. The implementation of the methodology in full core codes is discussed with particular attention to numerical issues and to the integration within the entire simulation process. Finally, first results from core studies in AN transport are analyzed in detail and compared to standard results of neutron diffusion.

  4. Composite Cores

    NASA Technical Reports Server (NTRS)

    1990-01-01

    Spang & Company's new configuration of converter transformer cores is a composite of gapped and ungapped cores assembled together in concentric relationship. The net effect of the composite design is to combine the protection from saturation offered by the gapped core with the lower magnetizing requirement of the ungapped core. The uncut core functions under normal operating conditions and the cut core takes over during abnormal operation to prevent power surges and their potentially destructive effect on transistors. Principal customers are aerospace and defense manufacturers. Cores also have applicability in commercial products where precise power regulation is required, as in the power supplies for large mainframe computers.

  5. The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel

    SciTech Connect

    David Petti; John Maki

    2005-02-01

    The fuel service conditions for the DOE Next Generation Nuclear Plant (NGNP) will be challenging. All major fuel related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO2 TRISO-coated particle fuel development program in the 1980s. While TRISO-coated particle fuel has been irradiated at NGNP relevant levels for two or three of the design parameters, no data exist for TRISO-coated particle fuel for all five parameters simultaneously. Of particular concern are the high burnup and high temperatures expected in the NGNP. In this paper, where possible, we evaluate the challenges associated with high burnup and high temperature quantitatively by examining the performance of the fuel in terms of different known failure mechanisms. Potential design solutions to ameliorate the negative effects of high burnup and high temperature are also discussed.

  6. Influence of the tritium beta(-) decay on low-temperature thermonuclear burn-up in deuterium-tritium mixtures

    PubMed

    Frolov

    2000-09-01

    Low-temperature (Tburn-up in deuterium-tritium mixtures with various deuterium-tritium-helium-3 ratios is considered. The general dependence is studied for the critical burn-up parameter x(c)=rhor(c) upon the initial temperature T, density rho(0), and tritium molar concentration y for the [D]:y[T]:(1-y)[3He] mixture. In particular, it is shown that, if the tritium concentration y decreases, then the critical burn-up parameter x(c)(T,rho(0),y) grows very quickly (at fixed T and rho(0)). This means that tritium beta(-) decay significantly complicates thermonuclear burn-up in deuterium-tritium mixtures.

  7. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    SciTech Connect

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.

  8. THE EFFECT OF BURNUP AND SEPARATION EFFICIENCY ON URANIUM UTILIZATION AND RADIOTOXICITY

    SciTech Connect

    Samuel Bays; Steven Piet

    2001-11-01

    This paper addresses two fundamental issues of fuel cycle sustainability. The two primary issues of interest are efficient use of the natural uranium resource (cradle), and management of nuclear waste radiotoxicity (grave). Both uranium utilization and radiotoxicity are directly influenced by the burnup achieved during irradiation (transmutation related) and where applicable the separation efficiency (partitioning related). Burnup influences the in-growth of transuranics by breeding them into the fuel cycle. Transuranic breeding is virtually essential to resource sustainability because it increases utilization of naturally abundant fertile U-238. However, the direct consequence of this build-up is the in-growth of transuranic isotopes which generally increase the source of future geologically committed radiotoxicity. For scenarios involving recycle, separation efficiency influences the degree to which this transuranic source term is removed from active service in the fuel stream and made a disposal legacy of human activity.

  9. Draft evaluation of the frequency for gas sampling for the high burnup confirmatory data project

    SciTech Connect

    Stockman, Christine T.; Alsaed, Halim A.; Bryan, Charles R.

    2015-03-26

    This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.

  10. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    SciTech Connect

    Kee, E.

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  11. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    SciTech Connect

    Burkes, Douglas E.; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-01

    A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U-Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U-Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U-Mo alloy decreased approximately 30% for a fission density of 2.88 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.08 × 1021 fissions cm-3 from unirradiated values at 200 oC. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  12. Experimental Test Plan for PWR Sister Rods in the High Burnup Spent Fuel Data Project

    SciTech Connect

    Montgomery, Rose; Scaglione, John M; Bevard, Bruce Balkcom; Hanson, Brady; Billone, Dr. Michael

    2016-01-01

    The High Burnup Spent Fuel Data project pulled 25 sister rods (9 from the project assemblies and 16 from similar HBU assemblies) for characterization. The 25 sister rods are all high burnup and cover the range of modern domestic cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of their long-term storage in a TN-32 dry storage cask (the cask rods ) to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup spent fuel.

  13. Method for adding additional isotopes to actinide-only burnup credit

    SciTech Connect

    Lancaster, D.B.; Fuentes, E.; Kang, C.

    1998-01-01

    The Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages requires computer code validation to be performed against a benchmark set of chemical assays for isotopic concentration and against a benchmark set of critical experiments for package criticality. Both sets contain all the isotopes included in the methodology. The chemical assays used include the uranium and plutonium isotopes, while the critical experiments were composed of UO{sub 2} or MOX rods, covering the isotopes in the actinide only approach. Since other isotopes are not included in the validation benchmark sets, it would be necessary to justify both the content and worth of any additional isotope for which burnup credit is to be taken (i.e., both the concentration and criticality effect of each particular isotope must be validated). A method is proposed here that can be used for any number of additional isotopes. As does the actinide-only burnup credit methodology, this method makes use of chemical assay data to establish the conservatism in the prediction of each isotope`s concentration. Criticality validation is also performed using a benchmark set of UO{sub 2} and MOX critical experiments, where the additional isotopes are validated using worth experiments to conservatively account for any uncertainty in their cross sections. The remaining requirements (analysis and modeling parameters, loading criteria generation, and physical implementation and controls) are performed exactly as described in the actinide-only burnup credit methodology. This report provides insight into each particular requirement in the new methodology.

  14. A Genesis breakup and burnup analysis in off-nominal Earth return and atmospheric entry

    NASA Technical Reports Server (NTRS)

    Salama, Ahmed; Ling, Lisa; McRonald, Angus

    2005-01-01

    The Genesis project conducted a detailed breakup/burnup analysis before the Earth return to determine if any spacecraft component could survive and reach the ground intact in case of an off-nominal entry. In addition, an independent JPL team was chartered with the responsibility of analyzing several definitive breakup scenarios to verify the official project analysis. This paper presents the analysis and results of this independent team.

  15. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

    SciTech Connect

    Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

    2006-10-31

    The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty,cycles (high burnup, boiling, aggressive chemistry) andto investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment.

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

    SciTech Connect

    Ilas, Germina; Gauld, Ian C

    2011-01-01

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

  17. Nuclide Composition Benchmark Data Set for Verifying Burnup Codes on Spent Light Water Reactor Fuels

    SciTech Connect

    Nakahara, Yoshinori; Suyama, Kenya; Inagawa, Jun; Nagaishi, Ryuji; Kurosawa, Setsumi; Kohno, Nobuaki; Onuki, Mamoru; Mochizuki, Hiroki

    2002-02-15

    To establish a nuclide composition benchmark data set for the verification of burnup codes, destructive analyses of light water reactor spent-fuel samples, which were cut out from several heights of spent-fuel rods, were carried out at the analytical laboratory at the Japan Atomic Energy Research Institute. The 16 samples from three kinds of pressurized water reactor (PWR) fuel rods and the 18 samples from two boiling water reactor (BWR) fuel rods were examined. Their initial {sup 235}U enrichments and burnups were from 2.6 to 4.1% and from 4 to 50 GWd/t, respectively. One PWR fuel rod and one BWR fuel rod contained gadolinia as a burnable poison. The measurements for more than 40 nuclides of uranium, transuranium, and fission product elements were performed by destructive analysis using mass spectrometry, and alpha-ray and gamma-ray spectrometry. Burnup for each sample was determined by the {sup 148}Nd method. The analytical methods and the results as well as the related irradiation condition data are compiled as a complete benchmark data set.

  18. Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel

    SciTech Connect

    Parks, C V; DeHart, M D; Wagner, John C

    2000-03-13

    This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan.

  19. Analysis of Burnup and Economic Potential of Alternative Fuel Materials in Thermal Reactors

    SciTech Connect

    Oggianu, Stella Maris; No, Hee Cheon; Kazimi, Mujid S.

    2003-09-15

    A strategy is proposed for the assessment of nuclear fuel material economic potential use in future light water reactors (LWRs). In this methodology, both the required enrichment and the fuel performance limits are considered. In order to select the best fuel candidate, the optimal burnup that produces the lowest annual fuel cost within the burnup potential for a given fuel material and smear density ratio is determined.Several nuclear materials are presented as examples of the application of the methodology proposed in this paper. The alternative fuels considered include uranium dioxide (UO{sub 2}), uranium carbide (UC), uranium nitride (UN), metallic uranium (U-Zr alloy), combined thorium and uranium oxides (ThO{sub 2}/UO{sub 2}), and combined thorium and uranium metals (U/Th). For these examples, a typical LWR lattice geometry in a zirconium-based cladding was assumed. The uncertainties in the results presented are large due to the scarcity of experimental data regarding the behavior of the considered materials at high burnups. Also, chemical compatibility issues are to be considered separately.The same methodology can be applied in the future to evaluate the economic potential of other nuclear fuel materials including different cladding designs, dispersions of ceramics into ceramics, dispersions of ceramics into metals, and also for geometries other than the traditional circular fuel pin.

  20. Calculating Path-Dependent Travel Time Prediction Variance and Covariance for the SALSA3D Global Tomographic P-Velocity Model with a Distributed Parallel Multi-Core Computer

    NASA Astrophysics Data System (ADS)

    Hipp, J. R.; Encarnacao, A.; Ballard, S.; Young, C. J.; Phillips, W. S.; Begnaud, M. L.

    2011-12-01

    Recently our combined SNL-LANL research team has succeeded in developing a global, seamless 3D tomographic P-velocity model (SALSA3D) that provides superior first P travel time predictions at both regional and teleseismic distances. However, given the variable data quality and uneven data sampling associated with this type of model, it is essential that there be a means to calculate high-quality estimates of the path-dependent variance and covariance associated with the predicted travel times of ray paths through the model. In this paper, we show a methodology for accomplishing this by exploiting the full model covariance matrix. Our model has on the order of 1/2 million nodes, so the challenge in calculating the covariance matrix is formidable: 0.9 TB storage for 1/2 of a symmetric matrix, necessitating an Out-Of-Core (OOC) blocked matrix solution technique. With our approach the tomography matrix (G which includes Tikhonov regularization terms) is multiplied by its transpose (GTG) and written in a blocked sub-matrix fashion. We employ a distributed parallel solution paradigm that solves for (GTG)-1 by assigning blocks to individual processing nodes for matrix decomposition update and scaling operations. We first find the Cholesky decomposition of GTG which is subsequently inverted. Next, we employ OOC matrix multiply methods to calculate the model covariance matrix from (GTG)-1 and an assumed data covariance matrix. Given the model covariance matrix we solve for the travel-time covariance associated with arbitrary ray-paths by integrating the model covariance along both ray paths. Setting the paths equal gives variance for that path. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.

  1. An alternative explanation for evidence that xenon depletion, pore formation, and grain subdivision begin at different local burnups

    NASA Astrophysics Data System (ADS)

    Rest, J.; Hofman, G. L.

    2000-01-01

    In order to interpret the recent observation that xenon depletion, pore formation, and grain subdivision occur successively at increasing local burnups, a rate-theory-based model is used to investigate the nucleation and growth of cavities during low-temperature irradiation of UO 2 in the presence of irradiation-induced interstitial-loop formation and growth. Consolidation of the dislocation structure takes into account the generation of forest dislocations and capture of interstitial dislocation loops. The loops accumulate and ultimately evolve into a low-energy cellular dislocation structure. The cell walls have been previously identified as recrystallization nuclei. The calculations indicate that nanometer-size bubbles are associated with this cellular dislocation structure while the observed micron-size bubbles are presumed to be either preexisting pores deformed by adjacent grains and/or new pores formed in the new recrystallized grain-boundary junctions. Subsequent to recrystallization, gas released from the recrystallized grains feeds the preexisting pores and the recrystallized grains may appear to form a preferential concentration of subdivided grains around the growing pores. This picture is illustrated in a sequence of photomicrographs of irradiated U 3O 8.

  2. Pre-conceptual design study of ASTRID core

    SciTech Connect

    Varaine, F.; Marsault, P.; Chenaud, M. S.; Bernardin, B.; Conti, A.; Sciora, P.; Venard, C.; Fontaine, B.; Devictor, N.; Martin, L.; Scholer, A. C.; Verrier, D.

    2012-07-01

    In the framework of the ASTRID project at CEA, core design studies are performed at CEA with the AREVA and EDF support. At the stage of the project, pre-conceptual design studies are conducted in accordance with GEN IV reactors criteria, in particularly for safety improvements. An improved safety for a sodium cooled reactor requires revisiting many aspects of the design and is a rather lengthy process in current design approach. Two types of cores are under evaluation, one classical derived from the SFR V2B and one more challenging called CFV (low void effect core) with a large gain on the sodium void effect. The SFR V2b core have the following specifications: a very low burn-up reactivity swing (due to a small cycle reactivity loss) and a reduced sodium void effect with regard to past designs such as the EFR (around 2$ minus). Its performances are an average burn-up of 100 GWd/t, and an internal conversion ratio equal to one given a very good behavior of this core during a control rod withdrawal transient). The CFV with its specific design offers a negative sodium void worth while maintaining core performances. In accordance of ASTRID needs for demonstration those cores are 1500 MWth power (600 MWe). This paper will focus on the CFV pre-conceptual design of the core and S/A, and the performances in terms of safety will be evaluated on different transient scenario like ULOF, in order to assess its intrinsic behavior compared to a more classical design like V2B core. The gap in term of margin to a severe accident due to a loss of flow initiator underlines the potential capability of this type of core to enhance prevention of severe accident in accordance to safety demonstration. (authors)

  3. Development and validation of burnup dependent computational schemes for the analysis of assemblies with advanced lattice codes

    NASA Astrophysics Data System (ADS)

    Ramamoorthy, Karthikeyan

    The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant

  4. RIA Limits Based On Commercial PWR Core Response To RIA

    SciTech Connect

    Beard, Charles L.; Mitchell, David B.; Slagle, William H.

    2006-07-01

    Reactivity insertion accident (RIA) limits have been under intense review by regulators since 1993 with respect to what should be the proper limit as a function of burnup. Some national regulators have imposed new lower limits while in the United States the limits are still under review. The data being evaluated with respect to RIA limits come from specialized test reactors. However, the use of test reactor data needs to be balanced against the response of a commercial PWR core in setting reasonable limits to insure the health and safety of the public without unnecessary restrictions on core design and operation. The energy deposition limits for a RIA were set in the 1970's based on testing in CDC (SPERT), TREAT, PBF and NSRR test reactors. The US limits given in radially averaged enthalpy are 170 cal/gm for fuel cladding failure and 280 cal/gm for coolability. Testing conducted in the 1990's in the CABRI, NSRR and IGR test reactors have demonstrated that the cladding failure threshold is reduced with burnup, with the primary impact due to hydrogen pickup for in-reactor corrosion. Based on a review of this data very low enthalpy limits have been proposed. In reviewing proposed limits from RIL-0401(1) it was observed that much of the data used to anchor the low allowable energy deposition levels was from recent NSRR tests which do not represent commercial PWR reactor conditions. The particular characteristics of the NSRR test compared to commercial PWR reactor characteristics are: - Short pulse width: 4.5 ms vs > 8 ms; - Low temperature conditions: < 100 deg. F vs 532 deg. F. - Low pressure environment: atmospheric vs {approx} 2200 psi. A review of the historical RIA database indicates that some of the key NSRR data used to support the RIL was atypical compared to the overall RIA database. Based on this detailed review of the RIA database and the response of commercial PWR core, the following view points are proposed. - The Failure limit should reflect local fuel

  5. Automated Design and Optimization of Pebble-bed Reactor Cores

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2010-07-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  6. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    NASA Astrophysics Data System (ADS)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  7. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    SciTech Connect

    Wagner, J. C.

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  8. Low Burnup Inert Matrix Fuels Performance: TRANSURANUS Analysis of the Halden IFA-652 First Irradiation Cycle

    SciTech Connect

    Calabrese, R.; Vettraino, F.; Tverberg, T.

    2006-07-01

    The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. A specific investigation on CSZ and thoria inert matrices has been developed by ENEA since several years. In-pile testing on the ENEA-conceived innovative fuels is ongoing in the OECD Halden HBWR since June 2000 (IFA-652 experiment). The registered burnup at the end of 2005 is about 38 MWd.kgU{sub eq}{sup -1} vs. 45 MWd.kgU{sub eq}{sup -1} (40 MWd.kgUOX{sub eq}{sup -1}) target. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. In this paper, the response at low burnup (< 7 MWd.kgU{sub eq}{sup -1}) of CSZ-based fuels loaded in IFA-652, is analysed by means of the TRANSURANUS code. To this purpose, a comprehensive modelling of the above mentioned un-irradiated fuels, mainly relying on the thermophysical characterisation performed at the JRC/ITU-Karlsruhe, has been implemented in a custom TRANSURANUS version (TU-IMF). A comparison of the code predictions vs. the experimental data, aimed at evaluating the early-stage under irradiation phenomena, particularly densification and relocation, has been performed. (authors)

  9. Determining Spent Nuclear Fuel's Plutonium Content, Initial Enrichment, Burnup, and Cooling Time

    SciTech Connect

    Cheatham, Jesse R; Francis, Matthew W

    2011-01-01

    The Next Generation of Safeguards Initiative is examining nondestructive assay techniques to determine the total plutonium content in spent nuclear fuel. The goal of this research was to develop new techniques that can independently verify the plutonium content in a spent fuel assembly without relying on an operator's declarations. Fundamentally this analysis sought to answer the following questions: (1) do spent fuel assemblies contain unique, identifiable isotopic characteristics as a function of their burnup, cooling time, and initial enrichment; (2) how much variation can be seen in spent fuel isotopics from similar and dissimilar reactor power operations; and (3) what isotopes (if any) could be used to determine burnup, cooling time, and initial enrichment? To answer these questions, 96,000 ORIGEN cases were run that simulated typical two-cycle operations with burnups ranging from 21,900 to 72,000 MWd/MTU, cooling times from 5 to 25 years, and initial enrichments between 3.5 and 5.0 weight percent. A relative error coefficient was determined to show how numerically close a reference solution has to be to another solution for the two results to be indistinguishable. By looking at the indistinguishable solutions, it can be shown how a precise measurement of spent fuel isotopics can be inconclusive when used in the absence of an operator's declarations. Using this Method of Indistinguishable Solutions (MIS), we evaluated a prominent method of nondestructive analysis - gamma spectroscopy. From this analysis, a new approach is proposed that demonstrates great independent forensic examination potential for spent nuclear fuel by examining both the neutron emissions of Cm-244 and the gamma emissions of Cs-134 and Eu-154.

  10. Development and validation of burnup dependent computational schemes for the analysis of assemblies with advanced lattice codes

    NASA Astrophysics Data System (ADS)

    Ramamoorthy, Karthikeyan

    The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant

  11. Core design studies for advanced burner test reactor.

    SciTech Connect

    Yang, W. S.; Kim, T. K.; Hill, R. N.; Nuclear Engineering Division

    2008-01-01

    TRU-based fuels. Preliminary design studies showed that it is feasible to design the ABTR to accommodate a wide range of conversion ratio (CR) by employing different assembly designs. The TRU enrichments required for various conversion ratios and the irradiation database suggested a phased approach with initial startup using conventional enrichment plutonium-based fuel and gradual transitioning to full core loading of transmutation fuel after its qualification phase (resulting in {approx}0.6 CR). The low CR transmutation fuel tests can be accommodated in the designated test assemblies, and if fully developed, core conversion to low CR fuel can be envisioned. Reference ABTR core designs with a rated power of 250 MWt were developed for ternary metal alloy and mixed oxide fuels based on WG-Pu feed. The reference core contains 54 driver, 6 test fuel, and 3 test material assemblies. For the startup core designs, the calculated TRU conversion ratio is 0.65 for the metal fuel core and 0.64 for the oxide fuel core. Both the metal and oxide cores show good performances. The metal fuel core requires an average TRU enrichment of 18.8% and yields a reactivity swing of 1.2 %{Delta}k over the 4-month cycle. The core average flux level is {approx}2.4 x 10{sup 15} n/cm{sup 2}s, and test assembly flux level is {approx}2.8 x 10{sup 15} n/cm{sup 2}s. Compared to the metal fuel core, the lower density oxide fuel core requires an average TRU enrichment of 21.8%, which results in a 780 kg TRU loading (as compared to 732 kg for metal) despite a {approx}9% smaller heavy metal inventory. The lower heavy metal inventory increases the burnup reactivity swing by {approx}10% and reduces the flux levels by {approx}8%. Alternative designs were also studied for a LWR-SF TRU feed and a low conversion ratio, including the recycle of the ABTR spent fuel TRU. The lower fissile contents of the LWR-SF TRU relative to the WG-Pu TRU significantly increase the required TRU enrichment of the startup cores to

  12. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    DOE PAGES

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; Sterbentz, James W.

    2014-09-03

    AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methodsmore » for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma

  13. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    SciTech Connect

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; Sterbentz, James W.

    2014-09-03

    AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry

  14. PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? EXTENDING CYCLE BURNUP

    SciTech Connect

    Primm, Trent; Chandler, David

    2009-01-01

    Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes that would be required to this procedure are identified. Design and safety related analyses that are required for the certification of a new fuel are identified. Qualification tests and comments regarding the regulatory approval process are provided along with a conceptual schedule.

  15. A Feasibility and Optimization Study to Design a Nondestructive ATR Fuel Permanent Scanning System to Determine Fuel Burnup

    NASA Astrophysics Data System (ADS)

    Navarro, J.; Ring, T. A.; Nigg, D. W.

    2014-04-01

    The goal of this project was to develop the best available non-destructive technique to determine burnup of the Advanced Test Reactor (ATR) fuels at Idaho National Laboratory, as well as to make a recommendation regarding the feasibility of implementing a permanent fuel scanning system at the ATR canal. The study determined that useful spectra for validation and fuel burnup predictions can be obtained in-situ at the ATR canal using three different detectors. In addition, the study established that calibration curves can be created to predict ATR fuel burnup onsite. The study also established that in order to design a rugged system that can stand the daily operations at the ATR canal a LaBr3 scintillator can be used effectively if deconvolution process is applied to increase the spectra resolution.

  16. Critical assessment of the pore size distribution in the rim region of high burnup UO2 fuels

    NASA Astrophysics Data System (ADS)

    Cappia, F.; Pizzocri, D.; Schubert, A.; Van Uffelen, P.; Paperini, G.; Pellottiero, D.; Macián-Juan, R.; Rondinella, V. V.

    2016-11-01

    A new methodology is introduced to analyse porosity data in the high burnup structure. Image analysis is coupled with the adaptive kernel density estimator to obtain a detailed characterisation of the pore size distribution, without a-priori assumption on the functional form of the distribution. Subsequently, stereological analysis is carried out. The method shows advantages compared to the classical approach based on the histogram in terms of detail in the description and accuracy within the experimental limits. Results are compared to the approximation of a log-normal distribution. In the investigated local burnup range (80-200 GWd/tHM), the agreement of the two approaches is satisfactory. From the obtained total pore density and mean pore diameter as a function of local burnup, pore coarsening is observed starting from ≈100 GWd/tHM, in agreement with a previous investigation.

  17. Determination of plutonium content in high burnup pressurized water reactor fuel samples and its use for isotope correlations for isotopic composition of plutonium.

    PubMed

    Joe, Kihsoo; Jeon, Young-Shin; Han, Sun-Ho; Lee, Chang-Heon; Ha, Yeong-Keong; Song, Kyuseok

    2012-06-01

    The content of plutonium isotopes in high burnup pressurized water reactor fuel samples was examined using both alpha spectrometry and mass spectrometry after anion exchange separation. The measured values were compared with results calculated by the ORIGEN-2 code. On average, the ratios (m/c) of the measured values (m) over the calculated values (c) were 1.22±0.16 for (238)Pu, 1.02±0.14 for (239)Pu, 1.08±0.06 for (240)Pu, 1.06±0.16 for (241)Pu, and 1.13±0.08 for (242)Pu. Using the Pu data obtained in this work, correlations were derived between the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), the alpha specific activities of Pu, and the atom % abundances of the Pu isotopes. Using these correlations, the atom % abundances of the plutonium isotopes in the target samples were calculated. These calculated results agreed within a range from 2 to 8% of the experimentally derived values according to the isotopes of plutonium.

  18. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses: Criticality (keff) Predictions

    SciTech Connect

    Scaglione, John M.; Mueller, Don E.; Wagner, John C.

    2014-12-01

    One of the most important remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation—in particular, the availability and use of applicable measured data to support validation, especially for fission products (FPs). Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. In this study, this paper describes a validation approach for commercial spent nuclear fuel (SNF) criticality safety (keff) evaluations based on best-available data and methods and applies the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The criticality validation approach utilizes not only available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion program to support validation of the principal actinides but also calculated sensitivities, nuclear data uncertainties, and limited available FP LCE data to predict and verify individual biases for relevant minor actinides and FPs. The results demonstrate that (a) sufficient critical experiment data exist to adequately validate keff calculations via conventional validation approaches for the primary actinides, (b) sensitivity-based critical experiment selection is more appropriate for generating accurate application model bias and uncertainty, and (c) calculated sensitivities and nuclear data uncertainties can be used for generating conservative estimates of bias for minor actinides and FPs. Results based on the SCALE 6.1 and the ENDF/B-VII.0 cross-section libraries indicate that a conservative estimate of the bias for the minor actinides and FPs is 1.5% of their worth

  19. Thermal Diffusivity Of Homogeneous SBR MOX Fuel With A Burn-up of 35 MWd/kgHM

    SciTech Connect

    Staicu, D.; Pagliosa, G.; Papaioannou, D.; Rondinella, V.V.; Cozzo, C.; Konings, R.; Walker, C.T.; Barker, M.; Weston, R.

    2007-07-01

    New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded 'laser-flash' device and show that the thermal diffusivity increases from the pellet periphery to the centre. Comparison shows that the thermal conductivity is in the same range than of UO{sub 2} of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of pile auto-irradiation. (authors)

  20. EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel

    SciTech Connect

    Teague, Melissa C.; Gorman, Brian P.; Miller, Brandon D.; King, Jeffrey

    2014-01-01

    Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to approximately 1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had approximately 2.5x higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice approximately 25 um cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel.

  1. Sodium fast reactor evaluation: Core materials

    NASA Astrophysics Data System (ADS)

    Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon; Raison, J. P.; Mizuno, T.; Delage, F.; Carmack, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

  2. Calculation of radiation induced swelling of uranium mononitride using the digital computer program CYGRO 2

    NASA Technical Reports Server (NTRS)

    Davison, H. W.; Fiero, I. B.

    1971-01-01

    Fuel volume swelling and clad diametral creep strains were calculated for five fuel pins, clad with either T-111 (Ta-8W-2.4Hf) or PWC-11 (Nb-1Zr-0.1C). The fuel pins were irradiated to burnups between 2.7 and 4.6%. Clad temperatures were between 1750 and 2400 F (1228 and 1589 K). The maximum percentage difference between calculated and experimentally measured values of volumetric fuel swelling is 60%.

  3. 24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES FOR A BRASS GATE VALVE BODY MADE ON A CORE BOX, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

  4. Modified Laser and Thermos cell calculations on microcomputers

    SciTech Connect

    Shapiro, A.; Huria, H.C.

    1987-01-01

    In the course of designing and operating nuclear reactors, many fuel pin cell calculations are required to obtain homogenized cell cross sections as a function of burnup. In the interest of convenience and cost, it would be very desirable to be able to make such calculations on microcomputers. In addition, such a microcomputer code would be very helpful for educational course work in reactor computations. To establish the feasibility of making detailed cell calculations on a microcomputer, a mainframe cell code was compiled and run on a microcomputer. The computer code Laser, originally written in Fortran IV for the IBM-7090 class of mainframe computers, is a cylindrical, one-dimensional, multigroup lattice cell program that includes burnup. It is based on the MUFT code for epithermal and fast group calculations, and Thermos for the thermal calculations. There are 50 fast and epithermal groups and 35 thermal groups. Resonances are calculated assuming a homogeneous system and then corrected for self-shielding, Dancoff, and Doppler by self-shielding factors. The Laser code was converted to run on a microcomputer. In addition, the Thermos portion of Laser was extracted and compiled separately to have available a stand alone thermal code.

  5. Metallic inert matrix fuel concept for minor actinides incineration to achieve ultra-high burn-up

    NASA Astrophysics Data System (ADS)

    Lipkina, K.; Savchenko, A.; Skupov, M.; Glushenkov, A.; Vatulin, A.; Uferov, O.; Ivanov, Y.; Kulakov, G.; Ershov, S.; Maranchak, S.; Kozlov, A.; Maynikov, E.; Konova, K.

    2014-09-01

    The advantages of using Inert Matrix Fuel (IMF) in a design of an isolated arrangement of fuel are considered, with emphasis on, low temperatures in the fuel center, achievement of high burn-ups, and an environment friendly process for the fuel element fabrication. Changes in the currently existing concept of IMF usage are suggested, involving novel IMF design in the nuclear fuel cycle.

  6. 77 FR 26050 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-02

    ... nuclear fuel (SNF) in transportation packages and storage casks. The draft SFST-ISG proposes to revise the... SNF systems, (d) add a recommendation for an optional misload analysis coupled with additional administrative SNF system loading procedures, in lieu of a direct burnup measurement, and (e) make...

  7. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail

    2010-12-01

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this

  8. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    SciTech Connect

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail,

    2010-12-23

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this

  9. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    SciTech Connect

    Trellue, Holly Renee; Fugate, Michael Lynn; Tobin, Stephen Joesph

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  10. Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

    SciTech Connect

    White, J.R.

    1980-08-01

    A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.

  11. Calculation of the radionuclides in PWR spent fuel samples for SFR experiment planning.

    SciTech Connect

    Naegeli, Robert Earl

    2004-06-01

    This report documents the calculation of radionuclide content in the pressurized water reactor (PWR) spent fuel samples planned for use in the Spent Fuel Ratio (SPR) Experiments at Sandia National Laboratories, Albuquerque, New Mexico (SNL) to aid in experiment planning. The calculation methods using the ORIGEN2 and ORIGEN-ARP computer codes and the input modeling of the planned PWR spent fuel from the H. B. Robinson and the Surry nuclear power plants are discussed. The safety hazards for the calculated nuclide inventories in the spent fuel samples are characterized by the potential airborne dose and by the portion of the nuclear facility hazard category 2 and 3 thresholds that the experiment samples would present. In addition, the gamma ray photon energy source for the nuclide inventories is tabulated to facilitate subsequent calculation of the direct and shielded dose rates expected from the samples. The relative hazards of the high burnup 72 gigawatt-day per metric ton of uranium (GWd/MTU) spent fuel from H. B. Robinson and the medium burnup 36 GWd/MTU spent fuel from Surry are compared against a parametric calculation of various fuel burnups to assess the potential for higher hazard PWR fuel samples.

  12. Geoantineutrino spectrum and slow nuclear burning on the boundary of the liquid and solid phases of the Earth's core

    NASA Astrophysics Data System (ADS)

    Rusov, V. D.; Pavlovich, V. N.; Vaschenko, V. N.; Tarasov, V. A.; Zelentsova, T. N.; Bolshakov, V. N.; Litvinov, D. A.; Kosenko, S. I.; Byegunova, O. A.

    2007-09-01

    We give an alternative description of the data produced in the KamLAND experiment. Assuming the existence of a natural nuclear reactor on the boundary of the liquid and solid phases of the Earth's core, a geoantineutrino spectrum is obtained. This assumption is based on the experimental results of V. Anisichkin and his collaborators on the interaction of uranium dioxide and uranium carbide with iron-nickel and silica-alumina melts at high pressure (5-10 GPa) and temperature (1600-2200°C), which led to the proposal of the existence of an actinide shell in the Earth's core. We describe the operating mechanism of this reactor as solitary waves of nuclear burning in 238U and/or 232Th medium, in particular, as neutron fission progressive waves of Feoktistov and/or Teller et al. type. Next, we propose a simplified model for the accumulation and burn-up kinetics in Feoktistov's U-Pu fuel cycle. We also apply this model for numerical simulations of neutron fission wave in a two-phase UO2/Fe medium on the surface of the Earth's solid core. The proposed georeactor model offers a mechanism for the generation of 3He. The 3He/4He distribution in the Earth's interior is calculated, which in turn can be used as a natural quantitative criterion of the georeactor thermal power. Finally, we give a tentative estimation of the geoantineutrino intensity and spectrum on the Earth's surface. For this purpose we use the O'Nions et al. geochemical model of mantle differentiation and crust growth complemented by a nuclear energy source (georeactor with power of 30 TW).

  13. Fission-detector determination of D-D triton burnup fraction in beam-heated TFTR (Tokamak Fusion Test Reactor) plasmas

    SciTech Connect

    Jassby, D.L.; Hendel, H.W.; Barnes, C.W.; Bosch, S.; Cecil, F.E.; McCune, D.C.; Nieschmidt, E.B.; Strachan, J.D.

    1987-06-01

    After the end of a neutral-beam injection pulse into a low-density TFTR plasma, once the beam-injected deuterons have thermalized, the neutron emission is dominated by the 14-MeV neutron production from D-D triton burnup. Ordinary fission detectors can measure the 14-MeV emission rate, which can be extrapolated back in time to estimate the equilibrium triton burnup fraction. The fractional burnup determined by this method is in the range of 0.3 to 1.5% for TFTR discharges to date, and is consistent with classical confinement and slowing down. 10 refs., 3 figs.

  14. Core layering

    NASA Astrophysics Data System (ADS)

    Jacobson, S. A.; Rubie, D. C.; Hernlund, J. W.; Morbidelli, A.

    2015-12-01

    We have created a planetary accretion and differentiation model that self-consistently builds and evolves Earth's core. From this model, we show that the core grows stably stratified as the result of rising metal-silicate equilibration temperatures and pressures, which increases the concentrations of light element impurities into each newer core addition. This stable stratification would naturally resist convection and frustrate the onset of a geodynamo, however, late giant impacts could mechanically mix the distinct accreted core layers creating large homogenous regions. Within these regions, a geodynamo may operate. From this model, we interpret the difference between the planetary magnetic fields of Earth and Venus as a difference in giant impact histories. Our planetary accretion model is a numerical N-body integration of the Grand Tack scenario [1]—the most successful terrestrial planet formation model to date [2,3]. Then, we take the accretion histories of Earth-like and Venus-like planets from this model and post-process the growth of each terrestrial planet according to a well-tested planetary differentiation model [4,5]. This model fits Earth's mantle by modifying the oxygen content of the pre-cursor planetesimals and embryos as well as the conditions of metal-silicate equilibration. Other non-volatile major, minor and trace elements included in the model are assumed to be in CI chondrite proportions. The results from this model across many simulated terrestrial planet growth histories are robust. If the kinetic energy delivered by larger impacts is neglected, the core of each planet grows with a strong stable stratification that would significantly impede convection. However, if giant impact mixing is very efficient or if the impact history delivers large impacts late, than the stable stratification can be removed. [1] Walsh et al. Nature 475 (2011) [2] O'Brien et al. Icarus 223 (2014) [3] Jacobson & Morbidelli PTRSA 372 (2014) [4] Rubie et al. EPSL 301

  15. Benchmark data for validating irradiated fuel compositions used in criticality calculations

    SciTech Connect

    Bierman, S.R.; Talbert, R.J.

    1994-10-01

    To establish criticality safety margins utilizing burnup credit in the storage and transport of spent reactor fuels requires a knowledge of the uncertainty in the calculated fuel composition used in making the reactivity assessment. To provide data for validating such calculated burnup fuel compositions, radiochemical assays have been obtained as part of the United States Department of Energy From-Reactor Cask Development Program. Assay results and associated operating histories on the initial three samples analyzed in this effort are presented. The three samples were taken from different axial regions of a Pressurized Water Reactor fuel rod and represent radiation exposures of about 37, 27, and 44 GWd/MTU. The data are presented in a benchmark type format to facilitate identification/referencing and computer code input.

  16. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    NASA Astrophysics Data System (ADS)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  17. Use of Burnup Credit as a Safety Factor in Handling of NIST Fuel Assemblies in the L Basin of SRS

    SciTech Connect

    Eghbali, DA

    2004-01-07

    Burnup credit was recently used for the first time in criticality safety analysis to support the handling of the National Institute of Standards and Technology spent fuel assemblies in the L Basin of Savannah River Site. Previous criticality safety analyses were based on the fissile content of fresh, unirradiated fuel assemblies, resulting in handling of a group of 10 or less fuel assemblies at a time. Using burnup credit, it was demonstrated that an isolated configuration of up to 14 NITS fuel assemblies, the maximum number of fuel assemblies in a full basket, submerged in a concrete-lined, water-filled pool is subcritical, resulting in several administrative controls being modified or eliminated without compromising safety.

  18. Viscosity of the earth's core.

    NASA Technical Reports Server (NTRS)

    Gans, R. F.

    1972-01-01

    Calculation of the viscosity of the core at the boundary of the inner and outer core. It is assumed that this boundary is a melting transition and the viscosity limits of the Andrade (1934,1952) hypothesis (3.7 to 18.5 cp) are adopted. The corresponding kinematic viscosities are such that the precessional system explored by Malkus (1968) would be unstable. Whether it would be sufficiently unstable to overcome a severely subadiabatic temperature gradient cannot be determined.

  19. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    SciTech Connect

    Asmolov, V.; Yegorova, L.

    1996-03-01

    Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.

  20. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    SciTech Connect

    Broadhead, B.L.

    1991-08-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications.

  1. Characterization of the non-uniqueness of used nuclear fuel burnup signatures through a Mesh-Adaptive Direct Search

    NASA Astrophysics Data System (ADS)

    Skutnik, Steven E.; Davis, David R.

    2016-05-01

    The use of passive gamma and neutron signatures from fission indicators is a common means of estimating used fuel burnup, enrichment, and cooling time. However, while characteristic fission product signatures such as 134Cs, 137Cs, 154Eu, and others are generally reliable estimators for used fuel burnup within the context where the assembly initial enrichment and the discharge time are known, in the absence of initial enrichment and/or cooling time information (such as when applying NDA measurements in a safeguards/verification context), these fission product indicators no longer yield a unique solution for assembly enrichment, burnup, and cooling time after discharge. Through the use of a new Mesh-Adaptive Direct Search (MADS) algorithm, it is possible to directly probe the shape of this "degeneracy space" characteristic of individual nuclides (and combinations thereof), both as a function of constrained parameters (such as the assembly irradiation history) and unconstrained parameters (e.g., the cooling time before measurement and the measurement precision for particular indicator nuclides). In doing so, this affords the identification of potential means of narrowing the uncertainty space of potential assembly enrichment, burnup, and cooling time combinations, thereby bounding estimates of assembly plutonium content. In particular, combinations of gamma-emitting nuclides with distinct half-lives (e.g., 134Cs with 137Cs and 154Eu) in conjunction with gross neutron counting (via 244Cm) are able to reasonably constrain the degeneracy space of possible solutions to a space small enough to perform useful discrimination and verification of fuel assemblies based on their irradiation history.

  2. MEMS Calculator

    National Institute of Standards and Technology Data Gateway

    SRD 166 MEMS Calculator (Web, free access)   This MEMS Calculator determines the following thin film properties from data taken with an optical interferometer or comparable instrument: a) residual strain from fixed-fixed beams, b) strain gradient from cantilevers, c) step heights or thicknesses from step-height test structures, and d) in-plane lengths or deflections. Then, residual stress and stress gradient calculations can be made after an optical vibrometer or comparable instrument is used to obtain Young's modulus from resonating cantilevers or fixed-fixed beams. In addition, wafer bond strength is determined from micro-chevron test structures using a material test machine.

  3. Chaotic Calculations.

    ERIC Educational Resources Information Center

    Chenery, Gordon

    1991-01-01

    Uses chaos theory to investigate the nonlinear phenomenon of population growth fluctuation. Illustrates the use of computers and computer programs to make calculations in a nonlinear difference equation system. (MDH)

  4. Neutronic and burnup characteristics of an actinide-reduced plutonium fuel with tungsten

    SciTech Connect

    Chang, G.S.

    1998-04-01

    The US and Russia expect to have a surplus of {approximately} 150 tonnes of weapons-grade plutonium (WGP) and 1,000 tonnes of weapons-grade uranium resulting from drastic reductions in nuclear weapons programs. One of the most favored candidate methods for disposing of the WGP is to blend it with natural or depleted uranium down to 5 to 7 wt% of WGP for light water reactor (LWR) fuel pellet fabrication. However, this approach, with a conversion ratio of 0.6, will produce more plutonium and other actinides in the spent fuel than the nonfertile fuel and the proposed actinide-reduced plutonium fuel (ARPF). This process only transforms the weapons-grade fissile materials to civilian-grade plutonium, which is still a non-proliferation concern, so it does not completely solve the plutonium disposition problem. Disposition of WGP in reactors without fertile material has been proposed by industry and national laboratories. A new ARPF is described that would use WGP mixed with medium-enrichment (20 at.% < {sup 235}U < 93 at.%) UO{sub 2} and the nonfertile material tungsten to achieve a conversion ratio < 0.1. The ARPF can meet the WGP disposal goal while minimizing the plutonium production. Its physics and burnup characteristics are analyzed, and the results are compared with LWR UO{sub 2} and mixed-oxide fuel.

  5. TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

    SciTech Connect

    Jian Gan; Brandon Miller; Dennis Keiser; Adam Robinson; James Madden; Pavel Medvedev; Daniel Wachs

    2014-04-01

    As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

  6. SEM Characterization of the High Burn-up Microstructure of U-7Mo Alloy

    SciTech Connect

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Jian Gan; Brandon Miller; Adam Robinson; Pavel Medvedev; James Madden; Dan Wachs; M. Teague

    2014-04-01

    During irradiation, the microstructure of U-7Mo evolves until at a fission density near 5x1021 f/cm3 a high-burnup microstructure exists that is very different than what was observed at lower fission densities. This microstructure is dominated by randomly distributed, relatively large, homogeneous fission gas bubbles. The bubble superlattice has collapsed in many microstructural regions, and the fuel grain sizes, in many areas, become sub-micron in diameter with both amorphous fuel and crystalline fuel present. Solid fission product precipitates can be found inside the fission gas bubbles. To generate more information about the characteristics of the high-fission density microstructure, three samples irradiated in the RERTR-7 experiment have been characterized using a scanning electron microscope equipped with a focused ion beam. The FIB was used to generate samples for SEM imaging and to perform 3D reconstruction of the microstructure, which can be used to look for evidence of possible fission gas bubble interlinkage.

  7. Mercury's Core

    NASA Astrophysics Data System (ADS)

    Peale, S. J.

    2005-05-01

    In determining Mercury's core structure from its rotational properties, the location of Cassini state 1 is crucial. Convincing radar evidence indicates that the mantle rests on a liquid layer (Margot et al. 2005), but there are no empirical constraints on the moment of inertia C/MR2, which constraints must wait for the determination of the gravitational coefficients J2 and C22 from the MESSENGER orbiting spacecraft, and an accurate determination of the obliquity of the Cassini state. Tidal and core-mantle dissipation drive the spin to the Cassini state with a time scale O(105) years, so the spin should occupy the Cassini state and thereby define its obliquity---unless there has been a recent excitation of a free precession of the spin. Another way the spin might be displaced from the Cassini state is if the variations in the orbital elements, which change the position of the Cassini state, cause the spin axis to lag behind as it attempts to follow the state. Fortunately, the solid angle the spin axis encloses as it precesses around the Cassini state is an adiabatic invariant, and it is conserved if the orbital element variations are slow compared to the precession rate. As the precession period is O(1000) years, and the time scales of orbital parameter variations are O(105) years, the spin axis should remain very close to the Cassini state if it were ever close. But how close is close? The increasing precision of the radar and eventual spacecraft measurements warrants a check on the likely proximity of the spin axis to the Cassini state. By numerically following the positions of the spin axis and Cassini state with orbital parameters varying with time scales and amplitudes comparable to the real variations, we show that the spin should remain within 1″ of the Cassini state once dissipative torques bring it there. The current spin axis position should thus define the Cassini state sufficiently to put reasonably tight constraints on the core structure

  8. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    NASA Astrophysics Data System (ADS)

    Navarro, Jorge

    The goal of this study presented is to determine the best available nondestructive technique necessary to collect validation data as well as to determine burnup and cooling time of the fuel elements on-site at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal, the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements nondestructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed were used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results, it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however, in order to enhance the quality of the spectra collected using this scintillator, a deconvolution method was developed. Following the development of the deconvolution method

  9. Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

    SciTech Connect

    Li, Zhian; Rahimi, Meraj; Tang, David; Aissa, Mourad; Flaganan, Michelle; Wagner, John C.

    2013-07-01

    The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup

  10. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    SciTech Connect

    Navarro, Jorge

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  11. Automated Core Design

    SciTech Connect

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2005-07-15

    Multistate searching methods are a subfield of distributed artificial intelligence that aims to provide both principles for construction of complex systems involving multiple states and mechanisms for coordination of independent agents' actions. This paper proposes a multistate searching algorithm with reinforcement learning for the automatic core design of a boiling water reactor. The characteristics of this algorithm are that the coupling structure and the coupling operation suitable for the assigned problem are assumed and an optimal solution is obtained by mutual interference in multistate transitions using multiagents. Calculations in an actual plant confirmed that the proposed algorithm increased the convergence ability of the optimization process.

  12. Gas core reactors for actinide transmutation and breeder applications

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1978-01-01

    This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

  13. Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files

    NASA Astrophysics Data System (ADS)

    Kozier, K. S.; Dyck, G. R.

    2006-04-01

    A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in alternate channels. Results were compared using different room-temperature data files for deuterium, various thermal-scattering-law data files for hydrogen bound in light water and deuterium bound in heavy water, and for pre-ENDF/B-VII and ENDF/B-VI.8 data for uranium. The reactivity differences observed were small (typically <1 mk) and increased with axial neutron leakage.

  14. HTTF Core Stress Analysis

    SciTech Connect

    Brian D. Hawkes; Richard Schultz

    2012-07-01

    In accordance with the need to determine whether cracking of the ceramic core disks which will be constructed and used in the High Temperature Test Facility (HTTF) for heatup and cooldown experiments, a set of calculation were performed using Abaqus to investigate the thermal stresses levels and likelihood for cracking. The calculations showed that using the material properties provided for the Greencast 94F ceramic, cracking is predicted to occur. However, this modeling does not predict the size or length of the actual cracks. It is quite likely that cracks will be narrow with rough walls which would impede the flow of coolant gases entering the cracks. Based on data recorded at Oregon State University using Greencast 94F samples that were heated and cooled at prescribed rates, it was concluded that the likelihood that the cracks would be detrimental to the experimental objectives is small.

  15. FY14 Status Report: CIRFT Testing Results on High Burnup UNF

    SciTech Connect

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

    2014-09-01

    The objective of this project is to perform a systematic study of SNF/UNF (spent nuclear fuel/or used nuclear fuel) integrity under simulated transportation environments by using hot cell testing technology developed recently at Oak Ridge National Laboratory (ORNL), CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester). Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmarking tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. With support from the US Department of Energy and the NRC, CIRFT testing has been continued. The CIRFT testing was conducted on three HBR rods (R3, R4, and R5), with two specimens failed and one specimen un-failed. The total number of cycles in the test of un-failed specimens went over 2.23 107; the test was stopped as because the specimen did not show any sign of failure. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of used fuel rods in terms of both the curvature amplitude and the maximum of absolute of curvature extremes. The latter is significant because the maxima of extremes signify the maximum of tensile stress of the outer fiber of the bending rod. So far, a large variety of hydrogen contents has been covered in the CIRFT testing on HBR rods. It has been shown that the load amplitude is the dominant factor that controls the lifetime of bending rods, but the hydrogen content also has an important effect on the lifetime attained, according to the load range tested.

  16. Start-up fuel and power flattening of sodium-cooled candle core

    SciTech Connect

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

    2013-07-01

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing.

  17. HOW STARLESS ARE STARLESS CORES?

    SciTech Connect

    Schnee, Scott; Friesen, Rachel; Di Francesco, James; Johnstone, Doug; Enoch, Melissa; Sadavoy, Sarah

    2012-01-20

    In this paper, we present the results of Combined Array for Research in Millimeter-wave Astronomy continuum and spectral line observations of the dense core Per-Bolo 45. Although this core has previously been classified as starless, we find evidence for an outflow and conclude that Per-Bolo 45 is actually an embedded, low-luminosity protostar. We discuss the impact of newly discovered, low-luminosity, embedded objects in the Perseus molecular cloud on starless core and protostar lifetimes. We estimate that the starless core lifetime has been overestimated by 4%-18% and the Class 0/I protostellar lifetime has been underestimated by 5%-20%. Given the relatively large systematic uncertainties involved in these calculations, variations on the order of 10% do not significantly change either core lifetimes or the expected protostellar luminosity function. Finally, we suggest that high-resolution (sub)millimeter surveys of known cores lacking near-infrared and mid-infrared emission are necessary to make an accurate census of starless cores.

  18. Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and Interim Storage of High-Burnup LWR Fuel

    SciTech Connect

    Gauld, I. C.; Ryman, J. C.

    2000-12-11

    This report investigates trends in the radiological decay properties and changes in relative nuclide importance associated with increasing enrichments and burnup for spent LWR fuel as they affect the areas of criticality safety, thermal analysis (decay heat), and shielding analysis of spent fuel transport and storage casks. To facilitate identifying the changes in the spent fuel compositions that most directly impact these application areas, the dominant nuclides in each area have been identified and ranked by importance. The importance is investigated as a function of increasing burnup to assist in identifying the key changes in spent fuel characteristics between conventional- and extended-burnup regimes. Studies involving both pressurized water-reactor (PWR) fuel assemblies and boiling-water-reactor (BWR) assemblies are included. This study is seen to be a necessary first step in identifying the high-burnup spent fuel characteristics that may adversely affect the accuracy of current computational methods and data, assess the potential impact on previous guidance on isotopic source terms and decay-heat values, and thus help identify areas for methods and data improvement. Finally, several recommendations on the direction of possible future code validation efforts for high-burnup spent fuel predictions are presented.

  19. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    DOE PAGES

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; et al

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  20. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    SciTech Connect

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; Fulwyler, James

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  1. Study of the triton-burnup process in different JET scenarios using neutron monitor based on CVD diamond

    NASA Astrophysics Data System (ADS)

    Nemtsev, G.; Amosov, V.; Meshchaninov, S.; Popovichev, S.; Rodionov, R.

    2016-11-01

    We present the results of analysis of triton burn-up process using the data from diamond detector. Neutron monitor based on CVD diamond was installed in JET torus hall close to the plasma center. We measure the part of 14 MeV neutrons in scenarios where plasma current varies in a range of 1-3 MA. In this experiment diamond neutron monitor was also able to detect strong gamma bursts produced by runaway electrons arising during the disruptions. We can conclude that CVD diamond detector will contribute to the study of fast particles confinement and help predict the disruption events in future tokamaks.

  2. WBGT Calculator

    SciTech Connect

    Hunter, Charles H.

    2000-05-22

    This software calculates a Wet Bulb Globe Temperature (WBGT) using standard measurements from a meteorological station. WBGT is used by Industrial Hygenists (IH) to determine heat stress potential to outdoor workers. Through the mid 1990''s, SRS technicians were dispatched several times daily to measure WBGT with a custom hand held instrument and results were dessiminated via telephone. Due to workforce reductions, the WSRC IH Department asked for the development of an automated method to simulate the WBGT measurement using existing real time data from the Atmospheric Technologies Group''s meteorological monitoring network.

  3. WBGT Calculator

    2000-05-22

    This software calculates a Wet Bulb Globe Temperature (WBGT) using standard measurements from a meteorological station. WBGT is used by Industrial Hygenists (IH) to determine heat stress potential to outdoor workers. Through the mid 1990''s, SRS technicians were dispatched several times daily to measure WBGT with a custom hand held instrument and results were dessiminated via telephone. Due to workforce reductions, the WSRC IH Department asked for the development of an automated method to simulatemore » the WBGT measurement using existing real time data from the Atmospheric Technologies Group''s meteorological monitoring network.« less

  4. Evolution of First Cores and Formation of Stellar Cores in Rotating Molecular Cloud Cores

    NASA Astrophysics Data System (ADS)

    Saigo, Kazuya; Tomisaka, Kohji; Matsumoto, Tomoaki

    2008-02-01

    We followed the collapse of cloud cores with various rotation speed and density frustrations using three-dimensional hydrodynamical simulations by assuming a barotropic equation of state and examined the comprehensive evolution paths from the rotation molecule cloud core to stellar core. We found that the evolutionary paths depend only on the angular velocity of initial cloud core Ωc0. These evolutionary paths agree well with predictions of Saigo and Tomisaka's quasi-equilibrium axisymmetric models and SPH calculations of Bate. Evolutionary paths are qualitatively classified into three types. (1) A slowly rotating cloud with Ωc0 < 0.01/tff = 0.05(ρc0/10-19 g cm -3)1/2 rad Myr -1 shows spherical-type evolution, where ρc0 is the initial central density. Such a cloud forms a first core which is mainly supported by the thermal pressure. The first core has a small mass of Mcore ~ 0.01 M⊙ and a short lifetime of a few ×100 yr. After exceeding the H2 dissociation density ρ simeq 5.6 × 10-8 g cm -3, it begins the second collapse, and the whole of the first core accretes onto the stellar core/disk within a few free-fall timescales. (2) A rotating cloud with 0.01/tff < Ωc0lesssim 0.05/tff shows disk-type evolution. In this case, the first core becomes a centrifugally supported massive disk with Mcore ~ a few × 0.01-0.1 M⊙ and the lifetime is a few thousand years. The first core is unstable against nonaxisymmetric dynamic instability and forms spiral arms. The gravitational torque through spiral structure extracts angular momentum from the central region to the outer region of the first core. And only a central part with r ~ 1 AU begins the second collapse after exceeding dissociation density. However, the outer remnant disk keeps its centrifugal balance after stellar core formation. It seems that this remnant of the first core should control the mass and angular momentum accretion onto the newborn stellar system. (3) A rotating cloud with 0.05/tfflesssim Ωc0

  5. Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems.

    1999-07-23

    AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energymore » deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations.« less

  6. Dual-core antiresonant hollow core fibers.

    PubMed

    Liu, Xuesong; Fan, Zhongwei; Shi, Zhaohui; Ma, Yunfeng; Yu, Jin; Zhang, Jing

    2016-07-25

    In this work, dual-core antiresonant hollow core fibers (AR-HCFs) are numerically demonstrated, based on our knowledge, for the first time. Two fiber structures are proposed. One is a composite of two single-core nested nodeless AR-HCFs, exhibiting low confinement loss and a circular mode profile in each core. The other has a relatively simple structure, with a whole elliptical outer jacket, presenting a uniform and wide transmission band. The modal couplings of the dual-core AR-HCFs rely on a unique mechanism that transfers power through the air. The core separation and the gap between the two cores influence the modal coupling strength. With proper designs, both of the dual-core fibers can have low phase birefringence and short modal coupling lengths of several centimeters.

  7. Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor

    SciTech Connect

    Khericha, S.T.

    2002-06-30

    This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to {approx}42 GWd/MT burnup (+ 2.5%) as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: {approx}50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies ({at} {approx}40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches {approx}40 GWd/MT burnup per MCNP-predicted values.

  8. Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor

    SciTech Connect

    Khericha, Soli T

    2002-06-01

    This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to ~42 GWd/MT burnup (+ 2.5% as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: ~50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies (@ ~40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches ~40 GWd/MT burnup per MCNP-predicted values.

  9. Inner Core Rotation from Geomagnetic Westward Drift and a Stationary Spherical Vortex in Earth's Core

    NASA Technical Reports Server (NTRS)

    Voorhies, C. V.

    1999-01-01

    The idea that geomagnetic westward drift indicates convective leveling of the planetary momentum gradient within Earth's core is pursued in search of a differentially rotating mean state, upon which various oscillations and secular effects might be superimposed. The desired state conforms to roughly spherical boundary conditions, minimizes dissipative interference with convective cooling in the bulk of the core, yet may aide core cooling by depositing heat in the uppermost core and lower mantle. The variational calculus of stationary dissipation applied to a spherical vortex within the core yields an interesting differential rotation profile akin to spherical Couette flow bounded by thin Hartmann layers. Four boundary conditions are required. To concentrate shear induced dissipation near the core-mantle boundary, these are taken to be: (i) no-slip at the core-mantle interface; (ii) geomagnetically estimated bulk westward flow at the base of the core-mantle boundary layer; (iii) no-slip at the inner-outer core interface; and, to describe magnetic locking of the inner core to the deep outer core, (iv) hydrodynamically stress-free at the inner-outer core boundary. By boldly assuming the axial core angular momentum anomaly to be zero, the super-rotation of the inner core is calculated to be at most 1.5 degrees per year.

  10. Inner Core Rotation from Geomagnetic Westward Drift and a Stationary Spherical Vortex in Earth's Core

    NASA Technical Reports Server (NTRS)

    Voorhies, Coerte V.

    1998-01-01

    The idea that geomagnetic westward drift indicates convective leveling of the planetary momentum gradient within Earth's core is pursued in search of a differentially rotating mean state, upon which various oscillations and secular effects might be superimposed. The desired state conforms to roughly spherical boundary conditions, minimizes dissipative interference with convective cooling in the bulk of the core, yet may aid core cooling by depositing heat in the uppermost core and lower mantle. The variational calculus of stationary dissipation applied to a spherical vortex within the core yields an interesting differential rotation profile, akin to spherical Couette flow bounded by thin Hartmann layers. Four boundary conditions are required. To concentrate shear induced dissipation near the core-mantle boundary, these are taken to be: (i) no-slip at the core-mantle interface; (ii) geomagnetically estimated bulk westward flow at the base of the core-mantle boundary layer; (iii) no-slip at the inner-outer core interface; and, to describe magnetic locking of the inner core to the deep outer core; (iv) hydrodynamically stress-free at the inner-outer core boundary. By boldly assuming the axial core angular momentum anomaly to be zero, the super-rotation of the inner core relative to the mantle is calculated to be at most 1.5 deg./yr.

  11. Solvent extraction studies with high-burnup Fast Flux Test Facility fuel in the Solvent Extraction Test Facility

    SciTech Connect

    Benker, D.E.; Bigelow, J.E.; Bond, W.D.; Chattin, F.R.; King, L.J.; Kitts, F.G.; Ross, R.G.; Stacy, R.G.

    1986-10-01

    A batch of high-burnup fuel from the Fast Flux Test Facility (FFTF) was processed in the Solvent Extraction Test Facility (SETF) during Campaign 9. The fuel had a burnup of {similar_to}0 MWd/kg and a cooling time of {similar_to} year. Two runs were made with this fuel; in the first, the solvent contained 30% tri-n-butyl phosphate (TBP) and partitioning of the uranium and plutonium was effected by reducing the plutonium with hydroxylamine nitrate (HAN); in the second, the solvent contained 10% TBP and a low operating temperature was used in an attempt to partition without reducing the plutonium valence. The plutonium reoxidation problem, which was present in previous runs that used HAN, may have been solved by lowering the temperature and acidity in the partition contactor. An automatic control system was used to maintain high loadings of heavy metals in the coextraction-coscrub contactor in order to increase its efficiency while maintaining low losses of uranium and plutonium to the aqueous raffinate. An in-line photometer system was used to measure the plutonium concentration in an intermediate extraction stage; and based on this data, a computer algorithm determined the appropriate adjustments in the addition rate of the extractant. The control system was successfully demonstrated in a preliminary run with purified uranium. However, a variety of equipment and start up problems prevented an extended demonstration from being accomplished during the runs with the FFTF fuel.

  12. Neutron Transport and Nuclear Burnup Analysis for the Laser Inertial Confinement Fusion-Fission Energy (LIFE) Engine

    SciTech Connect

    Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E

    2008-10-24

    Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.

  13. Enhancement of REBUS-3/DIF3D for whole-core neutronic analysis of prismatic very high temperature reactor (VHTR).

    SciTech Connect

    Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A.; Nuclear Engineering Division

    2006-10-13

    Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficient data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing

  14. Pressure Vessel Calculations for VVER-440 Reactors

    NASA Astrophysics Data System (ADS)

    Hordósy, G.; Hegyi, Gy.; Keresztúri, A.; Maráczy, Cs.; Temesvári, E.; Vértes, P.; Zsolnay, É.

    2003-06-01

    Monte Carlo calculations were performed for a selected cycle of the Paks NPP Unit II to test a computational model. In the model the source term was calculated by the core design code KARATE and the neutron transport calculations were performed by the MCNP. Different forms of the source specification were examined. The calculated results were compared with measurements and in most cases fairly good agreement was found.

  15. Application of perturbation theory to lattice calculations based on method of cyclic characteristics

    NASA Astrophysics Data System (ADS)

    Assawaroongruengchot, Monchai

    computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and keff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and keff-EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these techniques to the CVR

  16. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    SciTech Connect

    Díez, C.J.; Cabellos, O.; Martínez, J.S.

    2015-01-15

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

  17. Code System to Process WIMSD4 Interface Output Files and Generate Two-Group Data for Reactor Calculations.

    1992-12-03

    Version 00 The code processes the WIMS-D/4 binary output files for producing two-group microscopic cross sections and macroscopic lattice cell constants (zone and cell macroscopic cross sections, D, M, and K-infinity) in a more flexible format needed for reactor burnup codes like CITATION, for reactor dynamics codes like NADYP-W and for other reactor codes. The purpose of the WIMSCORE-ENEA code is to facilitate the automation of data transfer between the cell calculation code WIMS andmore » the diffusion-burnup codes. Use is made of the VARY storage manipulation package. WIMSCORE generates output files to be used by the codes TDB, TRITON, CITATION.« less

  18. The whole-core LEU silicide fuel demonstration in the JMTR

    SciTech Connect

    Aso, Tomokazu; Akashi, Kazutomo; Nagao, Yoshiharu

    1997-08-01

    The JMTR was fully converted to LEU silicide (U{sub 3}Si{sub 2}) fuel with cadmium wires as burnable absorber in January, 1994. The reduced enrichment program for the JMTR was initiated in 1979, and the conversion to MEU (enrichment ; 45%) aluminide fuel was carried out in 1986 as the first step of the program. The final goal of the program was terminated by the present LEU conversion. This paper describes the results of core physics measurement through the conversion phase from MEU fuel core to LEU fuel core. Measured excess reactivities of the LEU fuel cores are mostly in good agreement with predicted values. Reactivity effect and burnup of cadmium wires, therefore, were proved to be well predicted. Control rod worth in the LEU fuel core is mostly less than that in the MEU fuel core. Shutdown margin was verified to be within the safety limit. There is no significant difference in temperature coefficient of reactivity between the MEU and LEU fuel cores. These results verified that the JMTR was successfully and safely converted to LEU fuel. Extension of the operating cycle period was achieved and reduction of spend fuel elements is expected by using the fuel with high uranium density.

  19. Fully Coupled Modeling of Burnup-Dependent (U1- y , Pu y )O2- x Mixed Oxide Fast Reactor Fuel Performance

    NASA Astrophysics Data System (ADS)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

    During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increase of fuel burnup, which decreases the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect the fuel material properties, power distribution, and overall performance of the fuel pin. In order to investigate these important issues, the (U1- y Pu y )O2- x fuel pellet is studied by fully coupling thermal transport, deformation, oxygen diffusion, fission gas release and swelling, and plutonium redistribution to evaluate the effects on each other with burnup-dependent models, accounting for the evolution of fuel porosity. The approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with fast reactor mixed oxide fuel performance analysis. The modeling results showed a consistent fuel performance comparable with the previous results. Burnup degrades the fuel thermal conductivity, resulting in a significant fuel temperature increase. The fission gas release increased rapidly first and then steadily with the burnup increase. The fuel porosity increased dramatically at the beginning of the burnup and then kept constant as the fission gas released to the fuel free volume, causing the fuel temperature to increase. Another important finding is that the deviation from stoichiometry of oxygen affects greatly not only the fuel properties, for example, thermal conductivity, but also the fuel performance, for example, temperature distribution, porosity evolution, grain size growth, fission gas release, deformation, and plutonium redistribution. Special attention needs to be paid to the deviation from stoichiometry of oxygen in fuel fabrication. Plutonium content will also affect the fuel material properties and performance

  20. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    SciTech Connect

    Waeckel, N.; Cauvin, R.; Lebuffe, C.

    1997-01-01

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the {alpha}/{beta} transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates.

  1. JAEA Studies on High Burnup Fuel Behaviors during Reactivity-Initiated Accident and Loss-of-Coolant Accident

    SciTech Connect

    Fuketa, Toyoshi; Sugiyama, Tomoyuki; Nagase, Fumihisa; Suzuki, Motoe

    2007-07-01

    The objectives of fuel safety research program at Japan Atomic Energy Agency (JAEA) are; to evaluate adequacy of present safety criteria and safety margins; to provide a database for future regulation on higher burnup UO{sub 2} and MOX fuels, new cladding and pellets; and to provide reasonably mechanistic computer codes for regulatory application. The JAEA program is comprised of reactivity-initiated accident (RIA) studies including pulse-irradiation experiments in the NSRR and cladding mechanical tests, loss-of-coolant accident (LOCA) tests including integral thermal shock test and oxidation rate measurement, development and verification of computer codes FEMAXI-6 and RANNS, and so on. In addition to an overview of the fuel safety research at JAEA, most recent progresses in the RIA and LOCA tests programs and the codes development are described and discussed in the paper. (authors)

  2. Fission product release and microstructure changes during laboratory annealing of a very high burn-up fuel specimen

    NASA Astrophysics Data System (ADS)

    Hiernaut, J.-P.; Wiss, T.; Colle, J.-Y.; Thiele, H.; Walker, C. T.; Goll, W.; Konings, R. J. M.

    2008-07-01

    A commercial PWR fuel sample with a local burn-up of about 240 MWd/kgHM was annealed in a Knudsen cell mass spectrometer system with a heating rate of 10 K/min up to 2750 K at which temperature the sample was completely vaporized. The release of fission gases and fission products was studied as a function of temperature. In one of the runs the heating was interrupted successively at 900, 1500 and 1860 K and at each step a small fragment of the sample was examined by SEM and analysed by energy dispersive electron probe microanalysis. The release behaviour of volatile, gaseous and other less volatile fission products is presented and analysed with the EFFUS program and related to the structural changes of the fuel.

  3. Solvent extraction studies with intermediate-burnup Fast Flux Test Facility fuel in the Solvent Extraction Test Facility

    SciTech Connect

    Benker, D. E.; Bigelow, J. E.; Bond, W. D.; Chattin, F. R.; King, L. J.; Kitts, F. G.; Ross, R. G.; Stacy, R. G.

    1986-04-01

    In Campaign 8, two batches of irradiated fuel from the Fast Flux Test Facility (FFTF) were processed, using 30% TBP-NPH, in the Solvent Extraction Test Facility (SETF). The burnups were about 36 and 55 MWd/kg with 1.3- and 1-year cooling times, respectively. The latter fuel had the highest burnup and shortest cooling time of any fuel ever handled in the SETF. No major problems were noted during the operation of the mixer-settlers, and low uranium and plutonium losses (<0.02%) were achieved. Zirconium and ruthenium decontamination factors (DFs) were improved by increasing the number of scrub stages and increasing the peak solvent loading in the coextraction-coscrub bank. The use of an in-line photometer to measure the uranium and plutonium concentrations in a process stream permitted high solvent loadings of heavy metals to be achieved in the extraction bank while maintaining low losses to the aqueous raffinate. The investigation of two flowsheet options for making separate uranium and plutonium products (organic backscrub and selective uranium extraction) that was started in Campaign 7 was continued. High-quality products were again obtained (uranium and plutonium DFs of {similar_to}0{sup 4}). Plutonium reoxidation was still extensive even though hydrazine was added to the aqueous strip for the organic backscrub flowsheet. Two different plutonium oxalate precipitation procedures [Pu(III) and Pu(IV)] were used in the preparation of the plutonium oxide products; this was done so that the fuel fabrication characteristics of the oxide from the two procedures could be compared. A total of {similar_to}50 g of plutonium was recovered and shipped to the fuel refabrication program.

  4. Academic Rigor: The Core of the Core

    ERIC Educational Resources Information Center

    Brunner, Judy

    2013-01-01

    Some educators see the Common Core State Standards as reason for stress, most recognize the positive possibilities associated with them and are willing to make the professional commitment to implementing them so that academic rigor for all students will increase. But business leaders, parents, and the authors of the Common Core are not the only…

  5. Pressure vessel calculations for VVER-440 reactors.

    PubMed

    Hordósy, G; Hegyi, Gy; Keresztúri, A; Maráczy, Cs; Temesvári, E; Vértes, P; Zsolnay, E

    2005-01-01

    For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found.

  6. On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

    SciTech Connect

    Merk, B.; Weiss, F. P.

    2012-07-01

    The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

  7. Evolution of the core physics concept for the Canadian supercritical water reactor

    SciTech Connect

    Pencer, J.; Colton, A.; Wang, X.; Gaudet, M.; Hamilton, H.; Yetisir, M.

    2013-07-01

    The supercritical water cooled reactor (SCWR) is one of the advanced reactor concepts chosen by the GEN-IV International Forum (GIF) for research and development efforts. Canada's contribution is the Canadian SCWR, a heavy water moderated, pressure tube supercritical light water cooled reactor. Recent developments in the SCWR lattice and core concepts, primarily the introduction of a large central flow tube filled with coolant combined with a two-ring fuel assembly, have enabled significant improvements compared to earlier concepts. These improvements include a reduction in coolant void reactivity (CVR) by more than 10 mk, and an almost 40% increase in fuel exit burnup, which is achieved via balanced power distribution between the fuel pins in the fuel assembly. In this paper the evolution of the physics concept is reviewed, and the present lattice and core physics concepts are presented.

  8. Trends in adsorbate induced core level shifts

    NASA Astrophysics Data System (ADS)

    Nilsson, Viktor; Van den Bossche, Maxime; Hellman, Anders; Grönbeck, Henrik

    2015-10-01

    Photoelectron core level spectroscopy is commonly used to monitor atomic and molecular adsorption on metal surfaces. As changes in the electron binding energies are convoluted measures with different origins, calculations are often used to facilitate the decoding of experimental signatures. The interpretation could in this sense benefit from knowledge on trends in surface core level shifts for different metals and adsorbates. Here, density functional theory calculations have been used to systematically evaluate core level shifts for (111) and (100) surfaces of 3d, 4d, and 5d transition metals upon CO, H, O and S adsorption. The results reveal trends and several non-intuitive cases. Moreover, the difficulties correlating core level shifts with charging and d-band shifts are underlined.

  9. Core materials development for the fuel cycle R&D program

    SciTech Connect

    Maloy, S. A.; Toloczko, Mychailo B.; Cole, J. I.; Byun, Thak Sang

    2011-12-31

    The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels’ fast reactor core materials (cladding and duct) must be able to withstand very high doses (>300 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, radiation induced segregation, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI). To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350– 750 °C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510 °C. Compact tension, charpy and tensile specimens have been machined from this duct and mechanical testing as well as SANS and Mossbauer spectroscopy are currently being performed. Initial results from compression testing and Charpy testing reveal a strong increase in yield stress (~400 MPa) and a large increase in DBTT (up to 230 °C) for specimens irradiated at 383 °C to a dose of 28 dpa. Less hardening and a smaller increase in DBTT was observed for specimens irradiated at higher temperatures up to 500 °C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous

  10. Core materials development for the fuel cycle R&D program

    NASA Astrophysics Data System (ADS)

    Maloy, S. A.; Toloczko, M.; Cole, J.; Byun, T. S.

    2011-08-01

    The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels' fast reactor core materials (cladding and duct) must be able to withstand very high doses (>300 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, radiation induced segregation, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI). To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350-750 °C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510 °C. Compact tension, charpy and tensile specimens have been machined from this duct and mechanical testing as well as SANS and Mossbauer spectroscopy are currently being performed. Initial results from compression testing and Charpy testing reveal a strong increase in yield stress (˜400 MPa) and a large increase in DBTT (up to 230 °C) for specimens irradiated at 383 °C to a dose of 28 dpa. Less hardening and a smaller increase in DBTT was observed for specimens irradiated at higher temperatures up to 500 °C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous oxide

  11. Full core analysis of IRIS reactor by using MCNPX.

    PubMed

    Amin, E A; Bashter, I I; Hassan, Nabil M; Mustafa, S S

    2016-07-01

    This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code. PMID:27135607

  12. Full core analysis of IRIS reactor by using MCNPX.

    PubMed

    Amin, E A; Bashter, I I; Hassan, Nabil M; Mustafa, S S

    2016-07-01

    This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code.

  13. Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel

    SciTech Connect

    Eliseev, V.A.; Malisheva, I.V.; Matveev, V.I.; Egorov, A.V.; Maslov, P.A.

    2013-07-01

    With the development and research into the generation IV fast sodium reactors, great attention is paid to the enhancement of the core inherent self-protection characteristics. One of the problems dealt here is connected with the reduction of the reactivity margin so that the control rods running should not result in the core overheating and melting. In this paper we consider the possibilities of improving the core of BN-1200 with oxide fuel by a known method of introducing an axial fertile layer into the core. But unlike earlier studies this paper looks at the possibility of using such a layer not only for improving breeding, but also for reducing sodium void reactivity effect (SVRE). This proposed improvement of the BN-1200 core does not solve the problem of strong interference in control and protection system (CPS) rods of BN-1200, but they reduce significantly the reactivity margin for burn-up compensation. This helps compensate all the reactivity balances in the improved core configurations without violating constraints on SVRE value.

  14. Banded transformer cores

    NASA Technical Reports Server (NTRS)

    Mclyman, C. W. T. (Inventor)

    1974-01-01

    A banded transformer core formed by positioning a pair of mated, similar core halves on a supporting pedestal. The core halves are encircled with a strap, selectively applying tension whereby a compressive force is applied to the core edge for reducing the innate air gap. A dc magnetic field is employed in supporting the core halves during initial phases of the banding operation, while an ac magnetic field subsequently is employed for detecting dimension changes occurring in the air gaps as tension is applied to the strap.

  15. Mercury's thermal evolution and core crystallization regime

    NASA Astrophysics Data System (ADS)

    Rivoldini, A.; Van Hoolst, T.; Dumberry, M.; Steinle-Neumann, G.

    2015-10-01

    Unlike the Earth, where the liquid core isentrope is shallower than the core liquidus, at the lower pressures inside Mercury's core the isentrope can be steeper than the melting temperature. As a consequence, upon cooling, the isentrope may first enter a solid stability field near the core mantle boundary and produce ironrich snow that sinks under gravity and produces buoyant upwellings of iron depleted fluid. Similar to bottom up crystallization, crystallization initiated near the top might generate sufficient buoyancy flux to drive magnetic field generation by compositional convection.In this study we model Mercury's thermal evolution by taking into account the formation of iron-rich snow to assess when the conditions for an internally magnetic field can be satisfied. We employ a thermodynamic consistent description of the iron high-pressure phase diagram and thermoelastic properties of iron alloys as well as the most recent data about the thermal conductivity of core materials. We use a 1-dimensional parametrized thermal evolution model in the stagnant lid regime for the mantle (e.g. [1]) that is coupled to the core. The model for the mantle takes into account the formation of the crust due to melting at depth. Mantle convection is driven by heat producing radioactive elements, heat loss from secular cooling and from the heat supplied by the core. The heat generated inside the core is mainly provided from secular cooling, from the latent heat released at iron freezing, and from gravitational energy resulting form the release of light elements at the inner core-outer core boundary as well as from the sinking of iron-rich snow and subsequent upwellings of light elements in the snow zone. If the heat flow out of the core is smaller than the heat transported along the core isentrope a thermal boundary will from at the top of the outer core. To determine the extension of the convecting region inside the liquid core we calculate the convective power [2]. Finally, we

  16. Analysis of Fresh Fuel Critical Experiments Appropriate for Burnup Credit Validation

    SciTech Connect

    DeHart, M.D.

    1995-01-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

  17. HYDRATE CORE DRILLING TESTS

    SciTech Connect

    John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell

    2002-11-01

    The ''Methane Hydrate Production from Alaskan Permafrost'' project is a three-year endeavor being conducted by Maurer Technology Inc. (MTI), Noble, and Anadarko Petroleum, in partnership with the U.S. DOE National Energy Technology Laboratory (NETL). The project's goal is to build on previous and ongoing R&D in the area of onshore hydrate deposition. The project team plans to design and implement a program to safely and economically drill, core and produce gas from arctic hydrates. The current work scope includes drilling and coring one well on Anadarko leases in FY 2003 during the winter drilling season. A specially built on-site core analysis laboratory will be used to determine some of the physical characteristics of the hydrates and surrounding rock. Prior to going to the field, the project team designed and conducted a controlled series of coring tests for simulating coring of hydrate formations. A variety of equipment and procedures were tested and modified to develop a practical solution for this special application. This Topical Report summarizes these coring tests. A special facility was designed and installed at MTI's Drilling Research Center (DRC) in Houston and used to conduct coring tests. Equipment and procedures were tested by cutting cores from frozen mixtures of sand and water supported by casing and designed to simulate hydrate formations. Tests were conducted with chilled drilling fluids. Tests showed that frozen core can be washed out and reduced in size by the action of the drilling fluid. Washing of the core by the drilling fluid caused a reduction in core diameter, making core recovery very difficult (if not impossible). One successful solution was to drill the last 6 inches of core dry (without fluid circulation). These tests demonstrated that it will be difficult to capture core when drilling in permafrost or hydrates without implementing certain safeguards. Among the coring tests was a simulated hydrate formation comprised of coarse, large

  18. Energetics of core formation - A correction.

    NASA Technical Reports Server (NTRS)

    Flasar, F. M.; Birch, F.

    1973-01-01

    An error has recently been discovered in the calculation of the temperature rise conducted by Birch (1965) in connection with a determination of the release of gravitational energy accompanying a rapid formation of the earth's core from an initially undifferentiated state. A revised calculation of the energetic relations involved in the core formation is, therefore, presented. The radii of the initial undifferentiated cold earth and of the fictitious undifferentiated warm earth are both found to be slightly smaller than that of the present differentiated warm earth.

  19. On the consistent definition of spin-orbit effects calculated by relativistic effective core potentials with one-electron spin-orbit operators: Comparison of spin-orbit effects for Tl, TlH, TlH3, PbH2, and PbH4

    NASA Astrophysics Data System (ADS)

    Han, Young-Kyu; Bae, Cheolbeom; Lee, Yoon Sup

    1999-05-01

    The spin-orbit effects for Tl, TlH, TlH3, PbH2, and PbH4 are evaluated by two-component calculations using several relativistic effective core potentials (RECP) with one-electron spin-orbit operators. The used RECPs are shape-consistent RECPs derived by Wildman et al. [J. Chem. Phys. 107, 9975 (1997)] and three sets of energy-consistent (or adjusted) RECPs published by Schwerdtfeger et al. [Phys. Scr. 36, 453 (1987); J. Chem. Phys. 90, 762 (1989)], Küchle et al. [Mol. Phys. 74, 1245 (1991)], and Leininger et al. [Chem. Phys. 217, 19 (1997)]. The shape-consistent RECP results are in very good agreement with the Küchle et al. energy-consistent RECP results for all the molecules studied here and all-electron results for TlH. The RECPs of Schwerdtfeger et al. and Leininger et al. seem to provide qualitatively different spin-orbit effects. If one defines spin-free RECP as the potential average of the corresponding two-component RECP, all RECPs give very similar spin-orbit effects for all the cases. Most of the discrepancies of molecular spin-orbit effects among various RECPs reported in the literature may originate from different definitions of RECPs with or without a spin-orbit term and not from the inherent difference in spin-orbit operators.

  20. Disparity between beta-core levels in pregnancy urine and serum: implications for the origin of urinary beta-core.

    PubMed

    Wehmann, R E; Blithe, D L; Akar, A H; Nisula, B C

    1990-02-01

    We used a highly purified preparation of the naturally occurring core fragment of hCG beta (beta-core) and a new RIA for beta-core to investigate the concentrations and behavior of beta-core in serum and urine. We collected serum and 24-h urine specimens from healthy pregnant women during the first trimester of pregnancy. The concentrations of beta-core in serum were determined by analysis of fractions eluted from Sephadex G-100. Serum concentrations of beta-core immunoreactivity were very low (0.13-1.25 micrograms/L), while large amounts of beta-core were excreted in urine during pregnancy (as much as 4-5 mg/day). Interference with measurement by serum factors did not account for the low levels of beta-core immunoreactivity in pregnancy serum. Based on the known urinary clearance rate of beta-core in healthy nonpregnant subjects, we calculated that urinary clearance of serum beta-core accounts for only about 1% of the beta-core in pregnancy urine. We conclude that during pregnancy, the concentrations of beta-core in plasma are measurable, but extremely low, and that most of the beta-core in urine is derived by mechanisms other than urinary clearance of serum beta-core; most likely, these mechanisms involve nephrogenous production of beta-core from precursor molecules such as hCG and hCG beta.

  1. 23. CORE WORKER OPERATING A COREBLOWER THAT PNEUMATICALLY FILLED CORE ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    23. CORE WORKER OPERATING A CORE-BLOWER THAT PNEUMATICALLY FILLED CORE BOXES WITH RESIGN IMPREGNATED SAND AND CREATED A CORE THAT THEN REQUIRED BAKING, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

  2. Core-Cutoff Tool

    NASA Technical Reports Server (NTRS)

    Gheen, Darrell

    2007-01-01

    A tool makes a cut perpendicular to the cylindrical axis of a core hole at a predetermined depth to free the core at that depth. The tool does not damage the surrounding material from which the core was cut, and it operates within the core-hole kerf. Coring usually begins with use of a hole saw or a hollow cylindrical abrasive cutting tool to make an annular hole that leaves the core (sometimes called the plug ) in place. In this approach to coring as practiced heretofore, the core is removed forcibly in a manner chosen to shear the core, preferably at or near the greatest depth of the core hole. Unfortunately, such forcible removal often damages both the core and the surrounding material (see Figure 1). In an alternative prior approach, especially applicable to toxic or fragile material, a core is formed and freed by means of milling operations that generate much material waste. In contrast, the present tool eliminates the damage associated with the hole-saw approach and reduces the extent of milling operations (and, hence, reduces the waste) associated with the milling approach. The present tool (see Figure 2) includes an inner sleeve and an outer sleeve and resembles the hollow cylindrical tool used to cut the core hole. The sleeves are thin enough that this tool fits within the kerf of the core hole. The inner sleeve is attached to a shaft that, in turn, can be attached to a drill motor or handle for turning the tool. This tool also includes a cutting wire attached to the distal ends of both sleeves. The cutting wire is long enough that with sufficient relative rotation of the inner and outer sleeves, the wire can cut all the way to the center of the core. The tool is inserted in the kerf until its distal end is seated at the full depth. The inner sleeve is then turned. During turning, frictional drag on the outer core pulls the cutting wire into contact with the core. The cutting force of the wire against the core increases with the tension in the wire and

  3. HENRY'S LAW CALCULATOR

    EPA Science Inventory

    On-Site was developed to provide modelers and model reviewers with prepackaged tools ("calculators") for performing site assessment calculations. The philosophy behind OnSite is that the convenience of the prepackaged calculators helps provide consistency for simple calculations,...

  4. Raman micro-spectroscopy of UOX and MOX spent nuclear fuel characterization and oxidation resistance of the high burn-up structure

    NASA Astrophysics Data System (ADS)

    Jegou, C.; Gennisson, M.; Peuget, S.; Desgranges, L.; Guimbretière, G.; Magnin, M.; Talip, Z.; Simon, P.

    2015-03-01

    Raman micro-spectroscopy was applied to study the structure and oxidation resistance of UO2 (burnup 60 GWd/tHM) and MOX (burnup 47 GWd/tHM) irradiated fuels. The Raman technique, adapted to working under extreme conditions, enabled structural information to be obtained at the cubic micrometer scale in various zones of interest within irradiated fuel (central and zones like the Rim for UOX60, and the plutonium-enriched agglomerates for MOX47 characterized by a high burn-up structure), and the study of their oxidation resistance. As regards the structural information after irradiation, the spectra obtained make up a set of data consistent with the systematic presence of the T2g band characteristic of the fluorite structure, and of a triplet band located between 500 and 700 cm-1. The existence of this triplet can be attributed to the presence of defects originating in changes to the fuel chemistry occurring in the reactor (presence of fission products) and to the accumulation of irradiation damage. As concerns the oxidation resistance of the different zones of interest, Raman spectroscopy results confirmed the good stability of the restructured zones (plutonium-enriched agglomerates and Rim) rich in fission products compared to the non-restructured UO2 grains. A greater structural stability was noticed in the case of high plutonium content agglomerates, as this element favors the maintenance of the fluorite structure.

  5. Core sample extractor

    NASA Technical Reports Server (NTRS)

    Akins, James; Cobb, Billy; Hart, Steve; Leaptrotte, Jeff; Milhollin, James; Pernik, Mark

    1989-01-01

    The problem of retrieving and storing core samples from a hole drilled on the lunar surface is addressed. The total depth of the hole in question is 50 meters with a maximum diameter of 100 millimeters. The core sample itself has a diameter of 60 millimeters and will be two meters in length. It is therefore necessary to retrieve and store 25 core samples per hole. The design utilizes a control system that will stop the mechanism at a certain depth, a cam-linkage system that will fracture the core, and a storage system that will save and catalogue the cores to be extracted. The Rod Changer and Storage Design Group will provide the necessary tooling to get into the hole as well as to the core. The mechanical design for the cam-linkage system as well as the conceptual design of the storage device are described.

  6. The core paradox.

    NASA Technical Reports Server (NTRS)

    Kennedy, G. C.; Higgins, G. H.

    1973-01-01

    Rebuttal of suggestions from various critics attempting to provide an escape from the seeming paradox originated by Higgins and Kennedy's (1971) proposed possibility that the liquid in the outer core was thermally stably stratified and that this stratification might prove a powerful inhibitor to circulation of the outer core fluid of the kind postulated for the generation of the earth's magnetic field. These suggestions are examined and shown to provide no reasonable escape from the core paradox.

  7. Solvent extraction studies with low-burnup Fast Flux Test Facility fuel in the Solvent Extraction Test Facility

    SciTech Connect

    Benker, D.E.; Bigelow, J.E.; Bond, W.D.; Chattin, F.R.; King, L.J.; Kitts, F.G.; Ross, R.G.; Stacy, R.G.

    1985-01-01

    A batch of irradiated Fast Flux Test Facility (FFTF) fuel was processed for the first time in the Solvent Extraction Test Facility (SETF) at the Oak Ridge National Laboratory (ORNL) during Campaign 7. The average burnup of the fuel was only 0.2 atom %, but the cooling time was short enough ({similar_to}2 years) so that {sup 95}Zr was detected in the feed. This short cooling permitted our first measurement of {sup 95}Zr decontamination factors (DFs) without having to use tracers. No operational problems were noted in the operation of the extraction-scrubbing contactor, and low uranium and plutonium losses (< 0.01%) were measured. Fission product DFs were improved noticeably by increasing the number of scrub stages from six to eight. Two flowsheet options for making pure uranium and plutonium products (total partitioning) were tested. Each flowsheet used hydroxylamine nitrate for reducing plutonium. Good products were obtained (uranium DFs of > 10{sup 3} and plutonium DFs of > 10{sup 4}), but each flowsheet was troubled with plutonium reoxidation. Adding hydrazine and lowering the plutonium concentration lessened the problem but did not eliminate it. About 370 g of plutonium was recovered from these tests, purified by anion exchange, converted to PuO{sub 2}, and transferred to the fuel refabrication program. 7 references.

  8. Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

    NASA Astrophysics Data System (ADS)

    Uwaba, Tomoyuki; Ito, Masahiro; Maeda, Koji

    2011-09-01

    The C3M irradiation test, which was conducted in the experimental fast reactor, "Joyo", demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, "Monju". The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and 137Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

  9. Relationship between changes in the crystal lattice strain and thermal conductivity of high burnup UO 2 pellets

    NASA Astrophysics Data System (ADS)

    Amaya, Masaki; Nakamura, Jinichi; Fuketa, Toyoshi; Kosaka, Yuji

    2010-01-01

    Two kinds of disk-shaped UO 2 samples (4 mm in diameter and 1 mm in thickness) were irradiated in a test reactor up to about 60 and 130 GWd/t, respectively. The microstructures of the samples were investigated by means of optical microscopy, scanning electron microscopy/ electron probe micro-analysis (SEM/EPMA) and micro-X-ray diffractometry. The measured lattice parameters tended to be considerably smaller than the reported values, and the typical cauliflower structure which is often observed in high burnup fuel pellet is hardly seen in these samples. Thermal diffusivities of the samples were also measured by using a laser flash method, and their thermal conductivities were evaluated by multiplying the heat capacity of unirradiated UO 2 and sample densities. While the thermal conductivities of sample 2 showed recovery after being annealed at 1500 K, those of sample 4 were not clearly observed even after being annealed at 1500 K. These trends suggest that the amount of accumulated irradiation-induced defects depends on the irradiation condition of each sample. From the comparison of the changes in the lattice parameter and strain energy density before and after the thermal diffusivity measurements, it is likely that the thermal conductivity recovery in the temperature region from 1200 to 1500 K is related to the migration of dislocation.

  10. Alteration Behavior of High Burnup Spent Fuel in Salt Brine Under Hydrogen Overpressure and in Presence of Bromide

    SciTech Connect

    Loida, Andreas; Metz, Volker; Kienzler, Bernhard

    2007-07-01

    Recent studies have shown that in the presence of H2 overpressure, which forms due to the corrosion of the Fe based container, the dissolution rate of the spent fuel matrix is slowed down by a factor of about 10, associated with a distinct decrease of concentrations of important radionuclides. However, in a natural salt environment as well as in geological formations with chloride rich groundwater the presence of radiation chemically active impurities such as bromide must be taken in consideration. Bromide is known to react with {beta}/{gamma} radiolysis products, thus counteracting the protective H{sub 2} effect. In the present experiments using high burnup spent fuel, it is observed that during 212 days the matrix dissolution rate was enhanced by a factor of about 10 in the presence of up to 10{sup -3} M bromide and 3.2 bar H{sub 2} overpressure. However, concentrations of matrix bound actinides were found at the same level or below as found under identical conditions, but in the absence of bromide. In the long-term it is expected that the effect of bromide becomes less important, because the decrease of {beta}/{gamma}-activity results in a decrease of oxidative radicals, which react with bromide, while a-activity will dominate the radiation field. (authors)

  11. BWR AXIAL PROFILE

    SciTech Connect

    J. Huffer

    2004-09-28

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I.

  12. Core Research Center

    USGS Publications Warehouse

    Hicks, Joshua; Adrian, Betty

    2009-01-01

    The Core Research Center (CRC) of the U.S. Geological Survey (USGS), located at the Denver Federal Center in Lakewood, Colo., currently houses rock core from more than 8,500 boreholes representing about 1.7 million feet of rock core from 35 States and cuttings from 54,000 boreholes representing 238 million feet of drilling in 28 States. Although most of the boreholes are located in the Rocky Mountain region, the geologic and geographic diversity of samples have helped the CRC become one of the largest and most heavily used public core repositories in the United States. Many of the boreholes represented in the collection were drilled for energy and mineral exploration, and many of the cores and cuttings were donated to the CRC by private companies in these industries. Some cores and cuttings were collected by the USGS along with other government agencies. Approximately one-half of the cores are slabbed and photographed. More than 18,000 thin sections and a large volume of analytical data from the cores and cuttings are also accessible. A growing collection of digital images of the cores are also becoming available on the CRC Web site Internet http://geology.cr.usgs.gov/crc/.

  13. Core Forensics: Earth's Accretion and Differentiation

    NASA Astrophysics Data System (ADS)

    Badro, J.; Brodholt, J. P.; Siebert, J.; Piet, H.; Ryerson, F. J.

    2013-12-01

    Earth's accretion and its primitive differentiation are intimately interlinked processes. One way to constrain accretionary processes is by looking at the major differentiation event that took place during accretion: core formation. Understanding core formation and core composition can certainly shed a new light on early and late accretionary processes. On the other hand, testing certain accretionary models and hypothesis (fluxes, chemistries, timing) allows -short of validating them- at the very least to unambiguously refute them, through the 'filter'' of core formation and composition. Earth's core formed during accretion as a result of melting, phase-separation, and segregation of accretionary building blocks (from meteorites to planetesimals). The bulk composition of the core and mantle depends on the evolution (pressure, temperature, composition) of core extraction during accretion. The entire process left a compositional imprint on both reservoirs: (1) in the silicate Earth, in terms of siderophile trace-element (Ni, Co, V, Cr, among others) concentrations and isotopic fractionation (Si, Cu, among others), a record that is observed in present-day mantle rocks; and (2) on the core, in terms of major element composition and light elements dissolved in the metal, a record that is observed by seismology through the core density-deficit. This imprint constitutes actually a fairly impressive set of evidence (siderophile element concentration and fractionation, volatile and siderophile element isotopic fractionation), can be used today to trace back the primordial processes that occurred 4.5 billion years ago. We are seeking to provide an overhaul of the standard core formation/composition models, by using a new rationale that bridges geophysics and geochemistry. The new ingredients are (1) new laser-heated diamond anvil cell partitioning data, dramatically extending the previous P-T conditions for experimental work, (2) ab initio molecular dynamics calculations to

  14. Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

    SciTech Connect

    G. S. Chang; M. A. Lillo; R. G. Ambrosek

    2008-06-01

    The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the

  15. The compactness of presupernova stellar cores

    SciTech Connect

    Sukhbold, Tuguldur; Woosley, S. E.

    2014-03-01

    The success or failure of the neutrino-transport mechanism for producing a supernova in an evolved massive star is known to be sensitive not only to the mass of the iron core that collapses, but also to the density gradient in the silicon and oxygen shells surrounding that core. Here we study the systematics of a presupernova core's 'compactness' as a function of the mass of the star and the physics used in its calculation. Fine-meshed surveys of presupernova evolution are calculated for stars from 15 to 65 M {sub ☉}. The metallicity and the efficiency of semiconvection and overshoot mixing are both varied and bare carbon-oxygen cores are explored as well as full hydrogenic stars. Two different codes, KEPLER and MESA, are used for the study. A complex interplay of carbon and oxygen burning, especially in shells, can cause rapid variations in the compactness for stars of very nearly the same mass. On larger scales, the distribution of compactness with main sequence mass is found to be robustly non-monotonic, implying islands of 'explodabilty,' particularly around 8-20 M {sub ☉} and 25-30 M {sub ☉}. The carbon-oxygen (CO) core mass of a presupernova star is a better, (though still ambiguous) discriminant of its core structure than the main sequence mass.

  16. Can Psychiatric Rehabilitation Be Core to CORE?

    ERIC Educational Resources Information Center

    Olney, Marjorie F.; Gill, Kenneth J.

    2016-01-01

    Purpose: In this article, we seek to determine whether psychiatric rehabilitation principles and practices have been more fully incorporated into the Council on Rehabilitation Education (CORE) standards, the extent to which they are covered in four rehabilitation counseling "foundations" textbooks, and how they are reflected in the…

  17. Improvements of MCOR: A Monte Carlo depletion code system for fuel assembly reference calculations

    SciTech Connect

    Tippayakul, C.; Ivanov, K.; Misu, S.

    2006-07-01

    This paper presents the improvements of MCOR, a Monte Carlo depletion code system for fuel assembly reference calculations. The improvements of MCOR were initiated by the cooperation between the Penn State Univ. and AREVA NP to enhance the original Penn State Univ. MCOR version in order to be used as a new Monte Carlo depletion analysis tool. Essentially, a new depletion module using KORIGEN is utilized to replace the existing ORIGEN-S depletion module in MCOR. Furthermore, the online burnup cross section generation by the Monte Carlo calculation is implemented in the improved version instead of using the burnup cross section library pre-generated by a transport code. Other code features have also been added to make the new MCOR version easier to use. This paper, in addition, presents the result comparisons of the original and the improved MCOR versions against CASMO-4 and OCTOPUS. It was observed in the comparisons that there were quite significant improvements of the results in terms of k{sub inf}, fission rate distributions and isotopic contents. (authors)

  18. Hydrologic characterization of four cores from the Geysers Coring Project

    SciTech Connect

    Persoff, Peter; Hulen, Jeffrey B.

    1996-01-24

    Results of hydrologic tests conducted on four representative core plugs from Geysers Coring Project drill hole SB-15-D have been related to detailed mineralogic and textural characterization of the plugs to yield new information about permeability, porosity, and capillary-pressure characteristics of the uppermost Geysers steam reservoir and its immediately overlying caprock. The core plugs are all fine- to medium-grained, Franciscan-assemblage (late Mesozoic) metagraywacke with sparse Franciscan metamorphic quartz-calcite veins and late Cenozoic, hydrothermal quartz-calcite-pyrite veins. The matrices of three plugs from the caprock are rich in metamorphic mixed-layer illite/smectite and disseminated hydrothermal pyrite; the reservoir plug instead contains abundant illite and only minor pyrite. The reservoir plug and one caprock plug are sparsely disrupted by latest-stage, unmineralized microfractures which both follow and crosscut veinlets but which could be artifacts. Porosities of the plugs, measured by Boyles-law gas expansion, range between 1.9 and 2.5%. Gas permeability and Klinkenberg slip factor were calculated from gas-pressure-pulse-decay measurements using a specially designed permeameter with small (2 mL) reservoirs. Matrix permeabilities in the range 10-21 m² ( = 1 nanodarcy) were measured for two plugs that included mineral-filled veins but no unfilled microfractures. Greater permeabilities were measured on plugs that contained microfractures; at 500 psi net confining pressure, an effective aperture of 1.6 µm was estimated for one plug. Capillary pressure curves were determined for three cores by measuring saturation as weight gain of plugs equilibrated with atmospheres in which the relative humidity was controlled by saturated brines.

  19. Mercury's core evolution

    NASA Astrophysics Data System (ADS)

    Deproost, Marie-Hélène; Rivoldini, Attilio; Van Hoolst, Tim

    2016-10-01

    Remote sensing data of Mercury's surface by MESSENGER indicate that Mercury formed under reducing conditions. As a consequence, silicon is likely the main light element in the core together with a possible small fraction of sulfur. Compared to sulfur, which does almost not partition into solid iron at Mercury's core conditions and strongly decreases the melting temperature, silicon partitions almost equally well between solid and liquid iron and is not very effective at reducing the melting temperature of iron. Silicon as the major light element constituent instead of sulfur therefore implies a significantly higher core liquidus temperature and a decrease in the vigor of compositional convection generated by the release of light elements upon inner core formation.Due to the immiscibility in liquid Fe-Si-S at low pressure (below 15 GPa), the core might also not be homogeneous and consist of an inner S-poor Fe-Si core below a thinner Si-poor Fe-S layer. Here, we study the consequences of a silicon-rich core and the effect of the blanketing Fe-S layer on the thermal evolution of Mercury's core and on the generation of a magnetic field.

  20. Ice Core Investigations

    ERIC Educational Resources Information Center

    Krim, Jessica; Brody, Michael

    2008-01-01

    What can glaciers tell us about volcanoes and atmospheric conditions? How does this information relate to our understanding of climate change? Ice Core Investigations is an original and innovative activity that explores these types of questions. It brings together popular science issues such as research, climate change, ice core drilling, and air…

  1. NFE Core Bibliographies.

    ERIC Educational Resources Information Center

    Michigan State Univ., East Lansing. Inst. for International Studies in Education.

    This collection of core bibliographies, which expands on an initial bibliography published in 1979 of the core resources housed in the Non-Formal Education Information Center at Michigan State University, comprises a basic stock of materials on nonformal education and women in development that have been contributed by development planners,…

  2. CORE - Performance Feedback System

    SciTech Connect

    2009-10-02

    CORE is an architecture to bridge the gaps between disparate data integration and delivery of disparate information visualization. The CORE Technology Program includes a suite of tools and user-centered staff that can facilitate rapid delivery of a deployable integrated information to users.

  3. Iowa Core Annual Report

    ERIC Educational Resources Information Center

    Iowa Department of Education, 2015

    2015-01-01

    One central component of a great school system is a clear set of expectations, or standards, that educators help all students reach. In Iowa, that effort is known as the Iowa Core. The Iowa Core represents the statewide academic standards, which describe what students should know and be able to do in math, science, English language arts, and…

  4. Making an Ice Core.

    ERIC Educational Resources Information Center

    Kopaska-Merkel, David C.

    1995-01-01

    Explains an activity in which students construct a simulated ice core. Materials required include only a freezer, food coloring, a bottle, and water. This hands-on exercise demonstrates how a glacier is formed, how ice cores are studied, and the nature of precision and accuracy in measurement. Suitable for grades three through eight. (Author/PVD)

  5. Iron diffusion from first principles calculations

    NASA Astrophysics Data System (ADS)

    Wann, E.; Ammann, M. W.; Vocadlo, L.; Wood, I. G.; Lord, O. T.; Brodholt, J. P.; Dobson, D. P.

    2013-12-01

    The cores of Earth and other terrestrial planets are made up largely of iron1 and it is therefore very important to understand iron's physical properties. Chemical diffusion is one such property and is central to many processes, such as crystal growth, and viscosity. Debate still surrounds the explanation for the seismologically observed anisotropy of the inner core2, and hypotheses include convection3, anisotropic growth4 and dendritic growth5, all of which depend on diffusion. In addition to this, the main deformation mechanism at the inner-outer core boundary is believed to be diffusion creep6. It is clear, therefore, that to gain a comprehensive understanding of the core, a thorough understanding of diffusion is necessary. The extremely high pressures and temperatures of the Earth's core make experiments at these conditions a challenge. Low-temperature and low-pressure experimental data must be extrapolated across a very wide gap to reach the relevant conditions, resulting in very poorly constrained values for diffusivity and viscosity. In addition to these dangers of extrapolation, preliminary results show that magnetisation plays a major role in the activation energies for diffusion at low pressures therefore creating a break down in homologous scaling to high pressures. First principles calculations provide a means of investigating diffusivity at core conditions, have already been shown to be in very good agreement with experiments7, and will certainly provide a better estimate for diffusivity than extrapolation. Here, we present first principles simulations of self-diffusion in solid iron for the FCC, BCC and HCP structures at core conditions in addition to low-temperature and low-pressure calculations relevant to experimental data. 1. Birch, F. Density and composition of mantle and core. Journal of Geophysical Research 69, 4377-4388 (1964). 2. Irving, J. C. E. & Deuss, A. Hemispherical structure in inner core velocity anisotropy. Journal of Geophysical

  6. Internal core tightener

    DOEpatents

    Brynsvold, Glen V.; Snyder, Jr., Harold J.

    1976-06-22

    An internal core tightener which is a linear actuated (vertical actuation motion) expanding device utilizing a minimum of moving parts to perform the lateral tightening function. The key features are: (1) large contact areas to transmit loads during reactor operation; (2) actuation cam surfaces loaded only during clamping and unclamping operation; (3) separation of the parts and internal operation involved in the holding function from those involved in the actuation function; and (4) preloaded pads with compliant travel at each face of the hexagonal assembly at the two clamping planes to accommodate thermal expansion and irradiation induced swelling. The latter feature enables use of a "fixed" outer core boundary, and thus eliminates the uncertainty in gross core dimensions, and potential for rapid core reactivity changes as a result of core dimensional change.

  7. Lunar Core and Tides

    NASA Technical Reports Server (NTRS)

    Williams, J. G.; Boggs, D. H.; Ratcliff, J. T.

    2004-01-01

    Variations in rotation and orientation of the Moon are sensitive to solid-body tidal dissipation, dissipation due to relative motion at the fluid-core/solid-mantle boundary, and tidal Love number k2 [1,2]. There is weaker sensitivity to flattening of the core-mantle boundary (CMB) [2,3,4] and fluid core moment of inertia [1]. Accurate Lunar Laser Ranging (LLR) measurements of the distance from observatories on the Earth to four retroreflector arrays on the Moon are sensitive to lunar rotation and orientation variations and tidal displacements. Past solutions using the LLR data have given results for dissipation due to solid-body tides and fluid core [1] plus Love number [1-5]. Detection of CMB flattening, which in the past has been marginal but improving [3,4,5], now seems significant. Direct detection of the core moment has not yet been achieved.

  8. Mars' core and magnetism.

    PubMed

    Stevenson, D J

    2001-07-12

    The detection of strongly magnetized ancient crust on Mars is one of the most surprising outcomes of recent Mars exploration, and provides important insight about the history and nature of the martian core. The iron-rich core probably formed during the hot accretion of Mars approximately 4.5 billion years ago and subsequently cooled at a rate dictated by the overlying mantle. A core dynamo operated much like Earth's current dynamo, but was probably limited in duration to several hundred million years. The early demise of the dynamo could have arisen through a change in the cooling rate of the mantle, or even a switch in convective style that led to mantle heating. Presently, Mars probably has a liquid, conductive outer core and might have a solid inner core like Earth.

  9. Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

    NASA Astrophysics Data System (ADS)

    Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.

    2014-12-01

    Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

  10. Generally Contracted Valence-Core/Valence Basis Sets for Use with Relativistic Effective Core Potentials and Spin-Orbit Coupling Operators

    SciTech Connect

    Ermler, Walter V.; Tilson, Jeffrey L.

    2012-12-15

    A procedure for structuring generally contracted valence-core/valence basis sets of Gaussian-type functions for use with relativistic effective core potentials (gcv-c/v-RECP basis sets) is presented. Large valence basis sets are enhanced using a compact basis set derived for outer core electrons in the presence of small-core RECPs. When core electrons are represented by relativistic effective core potentials (RECPs), and appropriate levels of theory, these basis sets are shown to provide accurate representations of atomic and molecular valence and outer-core electrons. Core/valence polarization and correlation effects can be calculated using these basis sets through standard methods for treating electron correlation. Calculations of energies and spectra for Ru, Os, Ir, In and Cs are reported. Spectroscopic constants for RuO2+, OsO2+, Cs2 and InH are calculated and compared with experiment.

  11. PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

    NASA Astrophysics Data System (ADS)

    Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

    2014-06-01

    The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

  12. Dynamic origin of vortex core switching in soft magnetic nanodots.

    PubMed

    Guslienko, Konstantin Yu; Lee, Ki-Suk; Kim, Sang-Koog

    2008-01-18

    The magnetic vortex with in-plane curling magnetization and out-of-plane magnetization at the core is a unique ground state in nanoscale magnetic elements. This kind of magnetic vortex can be used, through its downward or upward core orientation, as a memory unit for information storage, and thus, controllable core switching deserves some special attention. Our analytical and micromagnetic calculations reveal that the origin of vortex core reversal is a gyrotropic field. This field is induced by vortex dynamic motion and is proportional to the velocity of the moving vortex. Our calculations elucidate the physical origin of the vortex core dynamic reversal, and, thereby, offer a key to effective manipulation of the vortex core orientation.

  13. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    SciTech Connect

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2010-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850ºC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 ºC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 to 30 MW, with various tests performed at each step to confirm

  14. 34. DESPATCH CORE OVENS, GREY IRON FOUNDRY CORE ROOM, BAKES ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    34. DESPATCH CORE OVENS, GREY IRON FOUNDRY CORE ROOM, BAKES CORES THAT ARE NOT MADE ON HEATED OR COLD BOX CORE MACHINES, TO SET BINDING AGENTS MIXED WITH THE SAND CREATING CORES HARD ENOUGH TO WITHSTAND THE FLOW OF MOLTEN IRON INSIDE A MOLD. - Stockham Pipe & Fittings Company, Grey Iron Foundry, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

  15. Enhanced transferability for Bethe-Salpeter Calculations

    NASA Astrophysics Data System (ADS)

    Shirley, Eric L.

    2015-03-01

    We have systematized projector-augmented-wave methods to reliably augment plane-wave/pseudopotential Bloch functions in atomic core regions for purposes of performing screening calculations, evaluating transition matrix elements, and evaluating Slater integrals in the condensed matter environment. This has improved the accuracy of core-hole screening, adherence to sum rules, and control of the strength of absorption features. This also ensures that transition matrix elements and concomitant core excitation spectra are reliable over significant energy ranges. To accomplish this, we improve the quality of the pseudopotentials (which become harder), extending norm conservation, and increasing the number of ``valence electrons.'' We present results for both insulators and metals, and for both core and valence excitations. Comparison to experimental data is a key part of this work. We also emphasize what approximations remain to be tackled in the treatment of electronic excitation spectra, many of which are more difficult to treat than what is within the scope of this work.

  16. Multiple Core Galaxies

    NASA Technical Reports Server (NTRS)

    Miller, R.H.; Morrison, David (Technical Monitor)

    1994-01-01

    Nuclei of galaxies often show complicated density structures and perplexing kinematic signatures. In the past we have reported numerical experiments indicating a natural tendency for galaxies to show nuclei offset with respect to nearby isophotes and for the nucleus to have a radial velocity different from the galaxy's systemic velocity. Other experiments show normal mode oscillations in galaxies with large amplitudes. These oscillations do not damp appreciably over a Hubble time. The common thread running through all these is that galaxies often show evidence of ringing, bouncing, or sloshing around in unexpected ways, even though they have not been disturbed by any external event. Recent observational evidence shows yet another phenomenon indicating the dynamical complexity of central regions of galaxies: multiple cores (M31, Markarian 315 and 463 for example). These systems can hardly be static. We noted long-lived multiple core systems in galaxies in numerical experiments some years ago, and we have more recently followed up with a series of experiments on multiple core galaxies, starting with two cores. The relevant parameters are the energy in the orbiting clumps, their relative.masses, the (local) strength of the potential well representing the parent galaxy, and the number of cores. We have studied the dependence of the merger rates and the nature of the final merger product on these parameters. Individual cores survive much longer in stronger background potentials. Cores can survive for a substantial fraction of a Hubble time if they travel on reasonable orbits.

  17. Separation of metallic residues from the dissolution of a high-burnup BWR fuel using nitrogen trifluoride

    SciTech Connect

    McNamara, Bruce K.; Buck, Edgar C.; Soderquist, Chuck Z.; Smith, Frances N.; Mausolf, Edward J.; Scheele, Randall D.

    2014-03-23

    Nitrogen trifluoride (NF3) was used to fluorinate the metallic residue from the dissolution of a high burnup, boiling water reactor fuel (~70 MWd/kgU). The metallic residue included the noble metal phase (containing ruthenium, rhodium, palladium, technetium, and molybdenum), and smaller amounts of zirconium, selenium, tellurium, and silver. Exposing the noble metal phase to 10% NF3 in argon between 400 and 550°C, removed molybdenum and technetium near 400°C as their volatile fluorides, and ruthenium near 500C as its volatile fluoride. The events were thermally and temporally distinct and the conditions specified are a recipe to separate these transition metals from each other and from the noble metal phase nonvolatile residue. Depletion of the volatile fluorides resulted in substantial exothermicity. Thermal excursion behavior was recorded under non-adiabatic, isothermal conditions that typically minimize heat release. Physical characterization of the metallic noble phase and its thermal behavior are consistent with high kinetic velocity reactions encouraged by the nanoparticulate phase or perhaps catalytic influences of the mixed platinum metals with nearly pure phase structure. Post-fluorination, only two phases were present in the residual nonvolatile fraction. These were identified as a nano-crystalline, metallic palladium cubic phase and a hexagonal rhodium trifluoride (RhF3) phase. The two phases were distinct as the sub-µm crystallites of metallic palladium were in contrast to the RhF3 phase, which grew from the parent nano-crystalline noble-metal phase during fluorination, to acicular crystals exceeding 20-µm in length.

  18. Advanced Fabrication Technique and Thermal Performance Prediction of U-Mo/Zr-alloy Dispersion Fuel Pin for High Burnup PWR

    NASA Astrophysics Data System (ADS)

    Suwardi

    2010-06-01

    In recent years, a novel class of zirconium alloys having the melting temperature of 990-1160 K has been developed. Based on novel zirconium matrix alloys, high uranium content fuel pin with U-9Mo has been developed according to capillary impregnation technique. The pin shows it is thermal conductivity ranging from 18 to 22 w/m/K that is comparably higher than UO2 pellet pin. The paper presents the met-met fabrication and thermal performance analysis of the fuel in typical PWR. The fabrication consists of mixing UO2 powder or granules and a novel Zr-alloy powder having low melting point, filling the mixture in a cladding tube that one of its end has been plugged, heating the pin to above melting temperature of Zr-alloy for an hour, natural cooling and heat treating at 300 K for 1/2 hr. The thermal analysis takes into account the pore and temperature distribution and high burn up effect to pellet conductivity. The thermal diffusivity ratio of novel to conventional fuel has been used as correction factor for the novel fuel conductivity. The results show a significant lowering pellet temperature along the radius until 1000 K at the hottest position. The analysis underestimates since the gap conductivity has been treated as decreased by 2% fission gas released that is not real since the use of lower temperature, and also decreasing thermal conductivity by porosity formation will much lower. The analysis shows that the novel fuel has very good thermal properties which able to pass the barrier of 65 MWD/kg-U, the limit to day commercial fuel. The burn-up extension means fewer fresh fuel is needed to produce electricity, preserve natural uranium resource, easier fuel handling operational per energy produced

  19. Global Core Plasma Model

    NASA Technical Reports Server (NTRS)

    Gallagher, Dennis L.; Craven, P. D.; Comfort, R. H.

    1999-01-01

    Abstract. The Global Core Plasma Model (GCPM) provides, empirically derived, core plasma density as a function of geomagnetic and solar conditions throughout the inner magnetosphere. It is continuous in value and gradient and is composed of separate models for the ionosphere, the plasmasphere, the plasmapause, the trough, and the polar cap. The relative composition of plasmaspheric H+, He+, and O+ is included in the GCPM. A blunt plasmaspheric bulge and rotation of the bulge with changing geomagnetic conditions is included. The GCPM is an amalgam of density models, intended to serve as a framework for continued improvement as new measurements become available and are used to characterize core plasma density, composition, and temperature.

  20. Core shroud corner joints

    DOEpatents

    Gilmore, Charles B.; Forsyth, David R.

    2013-09-10

    A core shroud is provided, which includes a number of planar members, a number of unitary corners, and a number of subassemblies each comprising a combination of the planar members and the unitary corners. Each unitary corner comprises a unitary extrusion including a first planar portion and a second planar portion disposed perpendicularly with respect to the first planar portion. At least one of the subassemblies comprises a plurality of the unitary corners disposed side-by-side in an alternating opposing relationship. A plurality of the subassemblies can be combined to form a quarter perimeter segment of the core shroud. Four quarter perimeter segments join together to form the core shroud.

  1. Distillation Calculations with a Programmable Calculator.

    ERIC Educational Resources Information Center

    Walker, Charles A.; Halpern, Bret L.

    1983-01-01

    Describes a three-step approach for teaching multicomponent distillation to undergraduates, emphasizing patterns of distribution as an aid to understanding the separation processes. Indicates that the second step can be carried out by programmable calculators. (A more complete set of programs for additional calculations is available from the…

  2. A NEW METHOD TO QUANTIFY CORE TEMPERATURE INSTABILITY IN RODENTS.

    EPA Science Inventory

    Methods to quantify instability of autonomic systems such as temperature regulation should be important in toxicant and drug safety studies. Stability of core temperature (Tc) in laboratory rodents is susceptible to a variety of stimuli. Calculating the temperature differential o...

  3. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  4. Modeling and design of a reload PWR core for a 48-month fuel cycle

    SciTech Connect

    McMahon, M.V.; Driscoll, M.J.; Todreas, N.E.

    1997-05-01

    The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7{sup w}/{sub o} U{sup 235} for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd{sub 2}O{sub 3}) mixed with the UO{sub 2} of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB{sub 2}) integral fuel burnable absorber (IFBA) coating on the Gd{sub 2}O{sub 3}-UO{sub 2} fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted.

  5. An innovative optical and chemical drill core scanner

    NASA Astrophysics Data System (ADS)

    Sjöqvist, A. S. L.; Arthursson, M.; Lundström, A.; Calderón Estrada, E.; Inerfeldt, A.; Lorenz, H.

    2015-05-01

    We describe a new innovative drill core scanner that semi-automatedly analyses drill cores directly in drill core trays with X-ray fluorescence spectrometry, without the need for much sample preparation or operator intervention. The instrument is fed with entire core trays, which are photographed at high resolution and scanned by a 3-D profiling laser. Algorithms recognise the geometry of the core tray, number of slots, location of the drill cores, calculate the optimal scanning path, and execute a continuous XRF analysis of 2 cm width along the core. The instrument is equipped with critical analytical components that allow an effective QA/QC routine to be implemented. It is a mobile instrument that can be manoeuvred by a single person with a manual pallet jack.

  6. Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges

    SciTech Connect

    B. McLeod

    2002-02-28

    This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M&O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M&O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report summarizes the

  7. Contaminated Sediment Core Profiling

    EPA Science Inventory

    Evaluating the environmental risk of sites containing contaminated sediments often poses major challenges due in part to the absence of detailed information available for a given location. Sediment core profiling is often utilized during preliminary environmental investigations ...

  8. Midland Core Repository

    SciTech Connect

    Tyler, Noel

    2000-08-14

    This report summarizes activities for this quarter in one table. Industrial users of this repository viewed and/or checked out 163 boxes of drill cores and cuttings samples from 18 wells during the quarter.

  9. Core helium flash

    SciTech Connect

    Cole, P.W.; Deupree, R.G.

    1980-01-01

    The role of convection in the core helium flash is simulated by two-dimensional eddies interacting with the thermonuclear runaway. These eddies are followed by the explicit solution of the 2D conservation laws with a 2D finite difference hydrodynamics code. Thus, no phenomenological theory of convection such as the local mixing length theory is required. The core helium flash is violent, producing a deflagration wave. This differs from the detonation wave (and subsequent disruption of the entire star) produced in previous spherically symmetric violent core helium flashes as the second dimension provides a degree of relief which allows the expansion wave to decouple itself from the burning front. Our results predict that a considerable amount of helium in the core will be burned before the horizontal branch is reached and that some envelope mass loss is likely.

  10. Biospecimen Core Resource - TCGA

    Cancer.gov

    The Cancer Genome Atlas (TCGA) Biospecimen Core Resource centralized laboratory reviews and processes blood and tissue samples and their associated data using optimized standard operating procedures for the entire TCGA Research Network.

  11. Core assembly storage structure

    DOEpatents

    Jones, Jr., Charles E.; Brunings, Jay E.

    1988-01-01

    A structure for the storage of core assemblies from a liquid metal-cooled nuclear reactor. The structure comprises an enclosed housing having a substantially flat horizontal top plate, a bottom plate and substantially vertical wall members extending therebetween. A plurality of thimble members extend downwardly through the top plate. Each thimble member is closed at its bottom end and has an open end adjacent said top plate. Each thimble member has a length and diameter greater than that of the core assembly to be stored therein. The housing is provided with an inlet duct for the admission of cooling air and an exhaust duct for the discharge of air therefrom, such that when hot core assemblies are placed in the thimbles, the heat generated will by convection cause air to flow from the inlet duct around the thimbles and out the exhaust duct maintaining the core assemblies at a safe temperature without the necessity of auxiliary powered cooling equipment.

  12. Azimuthal-spin-wave-mode-driven vortex-core reversals

    SciTech Connect

    Yoo, Myoung-Woo; Kim, Sang-Koog

    2015-01-14

    We studied, by micromagnetic numerical calculations, asymmetric vortex-core reversals driven by the m = −1 and m = +1 azimuthal spin-wave modes' excitations in soft magnetic circular nano-disks. We addressed the similarities and differences between the asymmetric core reversals in terms of the temporal evolutions of the correlated core-motion speed, locally concentrated perpendicular gyrofield, and magnetization dip near the original vortex core. The criterion for the core reversals was found to be the magnetization dip that must reach the out-of-plane magnetization component, m{sub z} = −p, with the initial polarization p, where p = +1 (−1) for the upward (downward) core magnetization. The core-motion speed and the associated perpendicular gyrofield, variable and controllable with static perpendicular field, H{sub z}, applied perpendicularly to the disk plane, must reach their threshold values to meet the ultimate core-reversal criterion. Also, we determined the H{sub z} strength and direction dependence of the core-switching time and threshold exciting field strength required for the core reversals, whose parameters are essential in the application aspect. This work offers deeper insights into the azimuthal spin-wave-driven core-reversal dynamics as well as an efficient means of controlling the azimuthal-modes-driven core reversals.

  13. A Technique to Determine Billet Core Charge Weight for P/M Fuel Tubes

    SciTech Connect

    Peacock, H.B.

    2001-07-02

    The core length in an extruded tube depends on the weight of powder in the billet core. In the past, the amount of aluminum powder needed to give a specified core length was determined empirically. This report gives a technique for calculating the weight of aluminum powder for the P/M core. An equation has been derived which can be used to determine the amount of aluminum needed for P/M billet core charge weights. Good agreement was obtained when compared to Mark 22 tube extrusion data. From the calculated charge weight, the elastomeric bag can be designed and made to compact the U3O8-Al core.

  14. Core-Noise Research

    NASA Technical Reports Server (NTRS)

    Hultgren, Lennart S.

    2012-01-01

    This presentation is a technical summary of and outlook for NASA-internal and NASA-sponsored external research on core noise funded by the Fundamental Aeronautics Program Subsonic Fixed Wing (SFW) Project. Sections of the presentation cover: the SFW system-level noise metrics for the 2015 (N+1), 2020 (N+2), and 2025 (N+3) timeframes; SFW strategic thrusts and technical challenges; SFW advanced subsystems that are broadly applicable to N+3 vehicle concepts, with an indication where further noise research is needed; the components of core noise (compressor, combustor and turbine noise) and a rationale for NASA's current emphasis on the combustor-noise component; the increase in the relative importance of core noise due to turbofan design trends; the need to understand and mitigate core-noise sources for high-efficiency small gas generators; and the current research activities in the core-noise area, with additional details given about forthcoming updates to NASA's Aircraft Noise Prediction Program (ANOPP) core-noise prediction capabilities, two NRA efforts (Honeywell International, Phoenix, AZ and University of Illinois at Urbana-Champaign, respectively) to improve the understanding of core-noise sources and noise propagation through the engine core, and an effort to develop oxide/oxide ceramic-matrix-composite (CMC) liners for broadband noise attenuation suitable for turbofan-core application. Core noise must be addressed to ensure that the N+3 noise goals are met. Focused, but long-term, core-noise research is carried out to enable the advanced high-efficiency small gas-generator subsystem, common to several N+3 conceptual designs, needed to meet NASA's technical challenges. Intermediate updates to prediction tools are implemented as the understanding of the source structure and engine-internal propagation effects is improved. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The

  15. Micro coring apparatus

    NASA Technical Reports Server (NTRS)

    Collins, David; Brooks, Marshall; Chen, Paul; Dwelle, Paul; Fischer, Ben

    1989-01-01

    A micro-coring apparatus for lunar exploration applications, that is compatible with the other components of the Walking Mobile Platform, was designed. The primary purpose of core sampling is to gain an understanding of the geological composition and properties of the prescribed environment. This procedure has been used extensively for Earth studies and in limited applications during lunar explorations. The corer is described and analyzed for effectiveness.

  16. Nuclear core positioning system

    DOEpatents

    Garkisch, Hans D.; Yant, Howard W.; Patterson, John F.

    1979-01-01

    A structural support system for the core of a nuclear reactor which achieves relatively restricted clearances at operating conditions and yet allows sufficient clearance between fuel assemblies at refueling temperatures. Axially displaced spacer pads having variable between pad spacing and a temperature compensated radial restraint system are utilized to maintain clearances between the fuel elements. The core support plates are constructed of metals specially chosen such that differential thermal expansion produces positive restraint at operating temperatures.

  17. Predicting Activation of Experiments Inside the Annular Core Research Reactor

    SciTech Connect

    Greenberg, Joseph Isaac

    2015-11-01

    The objective of this thesis is to create a program to quickly estimate the radioactivity and decay of experiments conducted inside of the Annular Core Research Reactor at Sandia National Laboratories and eliminate the need for users to write code. This is achieved by model the neutron fluxes in the reactor’s central cavity where experiments are conducted for 4 different neutron spectra using MCNP. The desired neutron spectrum, experiment material composition, and reactor power level are then input into CINDER2008 burnup code to obtain activation and decay information for every isotope generated. DREAD creates all of the files required for CINDER2008 through user selected inputs in a graphical user interface and executes the program for the user and displays the resulting estimation for dose rate at various distances. The DREAD program was validated by weighing and measuring various experiments in the different spectra and then collecting dose rate information after they were irradiated and comparing it to the dose rates that DREAD predicted. The program provides results with an average of 17% higher estimates than the actual values and takes seconds to execute.

  18. Autistic Savant Calendar Calculators.

    ERIC Educational Resources Information Center

    Patti, Paul J.

    This study identified 10 savants with developmental disabilities and an exceptional ability to calculate calendar dates. These "calendar calculators" were asked to demonstrate their abilities, and their strategies were analyzed. The study found that the ability to calculate dates into the past or future varied widely among these calculators. Three…

  19. MCNP LWR Core Generator

    SciTech Connect

    Fischer, Noah A.

    2012-08-14

    The reactor core input generator allows for MCNP input files to be tailored to design specifications and generated in seconds. Full reactor models can now easily be created by specifying a small set of parameters and generating an MCNP input for a full reactor core. Axial zoning of the core will allow for density variation in the fuel and moderator, with pin-by-pin fidelity, so that BWR cores can more accurately be modeled. LWR core work in progress: (1) Reflectivity option for specifying 1/4, 1/2, or full core simulation; (2) Axial zoning for moderator densities that vary with height; (3) Generating multiple types of assemblies for different fuel enrichments; and (4) Parameters for specifying BWR box walls. Fuel pin work in progress: (1) Radial and azimuthal zoning for generating further unique materials in fuel rods; (2) Options for specifying different types of fuel for MOX or multiple burn assemblies; (3) Additional options for replacing fuel rods with burnable poison rods; and (4) Control rod/blade modeling.

  20. Emergency core cooling system

    DOEpatents

    Schenewerk, William E.; Glasgow, Lyle E.

    1983-01-01

    A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.