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Sample records for burnup dependent group

  1. MTR core loading pattern optimization using burnup dependent group constants

    Directory of Open Access Journals (Sweden)

    Iqbal Masood

    2008-01-01

    Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.

  2. Burnup dependence of coolant void reactivity for ACR-1000 cell

    International Nuclear Information System (INIS)

    Le Tellier, R.; Marleau, G.; Hebert, A.; Roubstov, D.; Altiparmakov, D.; Irish, D.

    2008-01-01

    The Advanced Candu Reactor (ACR-1000) is light water cooled, fueled with enriched uranium and has a smaller lattice pitch than the Candu-6. As a result, the neutronic behavior of the ACR-1000 cell is expected to be somewhat different from that of the Candu-6 leading to a negative coolant void reactivity (CVR). Here we evaluate the CVR for the ACR-1000 cell using the lattice code DRAGON and compare our results with those obtained using the code WIMS-AECL. The differences observed between these two codes for the burnup dependence of the CVR is mainly explained in terms of the specific cell leakage model used by both codes. (authors)

  3. Burnup-dependent core neutronics analysis of plate-type research reactor using deterministic and stochastic methods

    International Nuclear Information System (INIS)

    Liu, Shichang; Wang, Guanbo; Liang, Jingang; Wu, Gaochen; Wang, Kan

    2015-01-01

    Highlights: • DRAGON & DONJON were applied in burnup calculations of plate-type research reactors. • Continuous-energy Monte Carlo burnup calculations by RMC were chosen as references. • Comparisons of keff, isotopic densities and power distribution were performed. • Reasons leading to discrepancies between two different approaches were analyzed. • DRAGON & DONJON is capable of burnup calculations with appropriate treatments. - Abstract: The burnup-dependent core neutronics analysis of the plate-type research reactors such as JRR-3M poses a challenge for traditional neutronics calculational tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity, large leakage and the particular neutron spectrum of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the burnup-dependent core neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON & DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic one. In the first stage, the homogenizations of few-group cross sections by DRAGON and the full core diffusion calculations by DONJON have been verified by comparing with the detailed Monte Carlo simulations. In the second stage, the burnup-dependent calculations of both assembly level and the full core level were carried out, to examine the capability of the deterministic code system DRAGON & DONJON to reliably simulate the burnup-dependent behavior of research reactors. The results indicate that both RMC and DRAGON & DONJON code system are capable of burnup-dependent neutronics analysis of research reactors, provided that appropriate treatments are applied in both assembly and core levels for the deterministic codes

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

    International Nuclear Information System (INIS)

    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 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 2 to higher oxides. The oxidation of UO 2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO 2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO 2 to UO 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 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 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 2 and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5)

  5. Improvements on burnup chain model and group cross section library in the SRAC system

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Okumura, Keisuke; Takano, Hideki; Ishiguro, Yukio; Kaneko, Kunio.

    1992-01-01

    Data and functions of the cell burnup calculation of the SRAC system were revised to improve mainly the accuracy of the burnup calculation of high conversion light water reactors (HCLWRs). New burnup chain models were developed in order to treat fission products (FPs) and actinide nuclides in detail. Group cross section library, SRACLIB-JENDL2, was generated based on JENDL-2 nuclear data file. In generating this library, emphasis was placed on FPs and actinides. Also revised were the data such as the average energy release per fission for various actinides. These improved data were verified by performing the burnup analysis of PWR spent fuels. Some new functions were added to the SRAC system for the convenience to yield macroscopic cross sections used in the core burnup process. (author)

  6. Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4

    International Nuclear Information System (INIS)

    Tombakoglu, M.; Cecen, Y.

    2001-01-01

    In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)

  7. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  8. VAMPIR - A two-group two-dimensional diffusion computer code for burnup calculation

    International Nuclear Information System (INIS)

    Zmijarevic, I.; Petrovic, I.

    1985-01-01

    VAMPIR is a computer code which simulates the burnup within a reactor coe. It computes the neutron flux, power distribution and burnup taking into account spatial variations of temperature and xenon poisoning. Its overall reactor calculation uses diffusion theory with finite differences approximation in X-Y or R-Z geometry. Two-group macroscopic cross section data are prepared by the lattice cell code WIMS-D4 and stored in the library form of multi entry tabulation against the various parameters that significantly affect the physical conditions in the reactor core. herein, the main features of the program are presented. (author)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

    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

  10. Dependence of heavy metal burnup on nuclear data libraries for fast reactors

    CERN Document Server

    Ohki, S

    2003-01-01

    Japan Nuclear Cycle Development Institute (JNC) is considering the highly burnt fuel as well as the recycling of minor actinide (MA) in the development of commercialized fast reactor cycle systems. Higher accuracy in burnup calculation is going to be required for higher mass plutonium isotopes ( sup 2 sup 4 sup 0 Pu, etc.) and MA nuclides. In the framework of research and development aiming at the validation and necessary improvements of fast reactor burnup calculation, we investigated the differences among the burnup calculation results with the major nuclear data libraries: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2, and JENDL-3.3. We focused on the heavy metal nuclides such as plutonium and MA in the central core region of a conventional sodium-cooled fast reactor. For main heavy metal nuclides ( sup 2 sup 3 sup 5 U, sup 2 sup 3 sup 8 U, sup 2 sup 3 sup 9 Pu, sup 2 sup 4 sup 0 Pu, and sup 2 sup 4 sup 1 Pu), number densities after 1-cycle burnup did not change over one or two percent. Library dependence was re...

  11. Reactivity effect of spent fuel depending on burn-up history

    International Nuclear Information System (INIS)

    Hayashi, Takafumi; Suyama, Kenya; Nomura, Yasushi

    2001-06-01

    It is well known that a composition of spent fuel depends on various parameter changes throughout a burn-up period. In this study we aimed at the boron concentration and its change, the coolant temperature and its spatial distribution, the specific power, the operation mode, and the duration of inspection, because the effects due to these parameters have not been analyzed in detail. The composition changes of spent fuel were calculated by using the burn-up code SWAT, when the parameters mentioned above varied in the range of actual variations. Moreover, to estimate the reactivity effect caused by the composition changes, the criticality calculations for an infinite array of spent fuel were carried out with computer codes SRAC95 or MVP. In this report the reactivity effects were arranged from the viewpoint of what parameters gave more positive reactivity effect. The results obtained through this study are useful to choose the burn-up calculation model when we take account of the burn-up credit in the spent fuel management. (author)

  12. Technique for sensitivity analysis of space- and energy-dependent burn-up calculations

    International Nuclear Information System (INIS)

    Williams, M.L.; White, J.R.

    1979-01-01

    A practical method is presented for sensitivity analysis of the very complex, space-energy dependent burn-up equations, in which the neutron and nuclide fields are coupled nonlinearly. The adjoint burn-up equations that are given are in a form which can be directly implemented into multi-dimensional depletion codes, such as VENTURE/BURNER. The data sensitivity coefficients can be used to determine the effect of data uncertainties on time-dependent depletion responses. Initial condition sensitivity coefficients provide a very effective method for computing the change in end of cycle parameters (such as k/sub eff/, fissile inventory, etc.) due to changes in nuclide concentrations at beginning of cycle

  13. Analysis of variation in few-group cross section behavior subjected to burnup and boron concentration

    International Nuclear Information System (INIS)

    Zhang Zongyao; Li Dongsheng.

    1986-01-01

    The paper analyzes the variations of few-group cross section behavior in neutron diffusion subjected to fuel burnup and critical boron concentration in a core. The influences of the behavior on the core excess reactivity, crirical boron concentration, power distribution and the yield of isotopes are also analyzed. A reactor core of samll-medium-sized nuclear power plant is analyzed as an example

  14. Grain size and burnup dependence of spent fuel oxidation: Geological repository impact

    International Nuclear Information System (INIS)

    Kansa, E.J.; Hanson, B.D.; Stout, R.B.

    1999-01-01

    Further refinements to the oxidation model of Stout et al. have been made. The present model incorporates the burnup dependence of the oxidation rate and an allowance for a distribution of grain sizes. The model was tested by comparing the model results with the oxidation histories of spent-fuel samples oxidized in thermogravimetric analysis (TGA) or oven dry-bath (ODB) experiments. The experimental and model results are remarkably close and confirm the assumption that grain-size distributions and activation energies are the important parameters to predicting oxidation behavior. The burnup dependence of the activation energy was shown to have a greater effect than decreasing the effective grain size in suppressing the rate of the reaction U 4 O 9 r↓U 3 O 8 . Model results predict that U 3 O 8 formation of spent fuels exposed to oxygen will be suppressed even for high burnup fuels that have undergone restructuring in the rim region, provided the repository temperature is kept sufficiently low

  15. Grain and burnup dependence of spent fuel oxidation: geological repository impact

    International Nuclear Information System (INIS)

    Hanson, B. D.; Kansa, E. J.; Stoot, R.B.

    1998-01-01

    Further refinements to the oxidation model of Stout et al. have been made. The present model incorporates the burnup dependence of the oxidation rate in addition to an allowance for a distribution of grain sizes. The model was tested by comparing the model results with the oxidation histories of spent fuel samples oxidized in Thermogravimetric Analysis (TGA) or Oven Dry-Bath (ODB) experiments. The comparison between the experimental and model results are remarkably close and confirm the assumption that grain-size distributions and activation energies are the important parameters to predicting oxidation behavior. The burnup dependence of the activation energy was shown to have a greater effect than decreasing the effective grain size in suppressing the rate of the reaction U 4 O 9 (rightwards arrow)U 3 O 4 . Model results predict that U 3 O 8 formation of spent fuels exposed to oxygen will be suppressed even for high burnup fuels that have undergone restructuring in the rim region, provided the repository temperature is kept sufficient

  16. ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for Candu Reactor Fuels

    International Nuclear Information System (INIS)

    2002-01-01

    and fission cross sections were obtained from collapsed 89-group ENDF/B-V and -VI data from the WIMS-AECL lattice code. Other reaction cross sections were obtained from the SCALE 27-group ENDF/B-IV data. In all, cross sections for more than 200 important actinides and fission products were updated with burnup-dependent data. The source of the nuclear decay data and cross sections not updated were the base ORIGEN-S libraries distributed with the SCALE 4.2 code package. The burnup-dependent cross sections were generated using eight burnup intervals that extend up to a burnup of approximately 12000 MWd/MgU. The cross sections for each interval are stored in positions on the ORIGEN-S library and are accessed by referencing the position number. The burnup values associated with each position are listed below. They are the same for both the 37-element and 28-element libraries. Library Position → Burnup (MWd/MgU): 1 → 240; 2 → 720; 3 → 1440; 4 → 2880; 5 → 4800; 6 → 6720; 7 → 8640; 8 → 10560

  17. Spatially dependent burnup implementation into the nodal program based on the finite element response matrix method

    International Nuclear Information System (INIS)

    Yoriyaz, H.

    1986-01-01

    In this work a spatial burnup scheme and feedback effects has been implemented into the FERM ( 'Finite Element Response Matrix' )program. The spatially dependent neutronic parameters have been considered in three levels: zonewise calculation, assembly wise calculation and pointwise calculation. Flux and power distributions and the multiplication factor were calculated and compared with the results obtained by CITATIOn program. These comparisons showed that processing time in the Ferm code has been hundred of times shorter and no significant difference has been observed in the assembly average power distribution. (Author) [pt

  18. OECD-NEA criticality working group - a status report and the burnup credit challenge

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1993-01-01

    A Working Group established by the organization for Economic Co-operation and Development's Nuclear Energy Agency (OECD-NEA), Paris, has examined the validity of computational methods used for calculations that evaluate the nuclear criticality safety issues involved in the storage, handling and transportation of fissile materials. The basic goal of this Working Group is to attempt to define and implement a procedure that can be shown to demonstrate the validity of the various computational methods used to make criticality safety calculations. The current activities of the Working Group involve an effort to establish the validity of computational methods used to evaluate the criticality safety of the storage, handling, and transportation of spent light-water-reactor fuel elements in which one seeks to take credit for the fissile material burnup and/or buildup of fission products. (J.P.N.)

  19. Implementation of burnup credit in spent fuel management systems. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1998-04-01

    The criticality safety analysis of spent fuel systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. Improved calculational methods allows one to take credit for the reactivity reduction associated with fuel burnup, hence reducing the analysis conservatism while maintaining an adequate criticality safety margin. Motivation for using burnup credit in criticality safety applications is generally based on economic considerations. Although economics may be a primary factor in deciding to use burnup credit, other benefits may be realized. Many of the additional benefits of burnup credit that are not strictly economic, may be considered to contribute to public health and safety, and resource conservation and environmental quality. Interest in the implementation of burnup credit has been shown by many countries. A summary of the information gathered by the IAEA about ongoing activities and regulatory status of burnup credit in different countries is included. Burnup credit implementation introduces new parameters and effects that should be addressed in the criticality analysis (e.g., axial and radial burnup shapes, fuel irradiation history, and others). Analysis of these parameters introduces new variations as well as the uncertainties, that should be considered in the safety assessment of the system. Also, the need arises to validate the isotopic composition that results from a depletion calculation, as well as to extend the current validation range of criticality codes to cover spent fuel. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Methods and procedures used in different countries are described in this report

  20. Implementation of burnup credit in spent fuel management systems. Proceedings of an advisory group meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-01

    The criticality safety analysis of spent fuel systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system`s reactivity. Improved calculational methods allows one to take credit for the reactivity reduction associated with fuel burnup, hence reducing the analysis conservatism while maintaining an adequate criticality safety margin. Motivation for using burnup credit in criticality safety applications is generally based on economic considerations. Although economics may be a primary factor in deciding to use burnup credit, other benefits may be realized. Many of the additional benefits of burnup credit that are not strictly economic, may be considered to contribute to public health and safety, and resource conservation and environmental quality. Interest in the implementation of burnup credit has been shown by many countries. A summary of the information gathered by the IAEA about ongoing activities and regulatory status of burnup credit in different countries is included. Burnup credit implementation introduces new parameters and effects that should be addressed in the criticality analysis (e.g., axial and radial burnup shapes, fuel irradiation history, and others). Analysis of these parameters introduces new variations as well as the uncertainties, that should be considered in the safety assessment of the system. Also, the need arises to validate the isotopic composition that results from a depletion calculation, as well as to extend the current validation range of criticality codes to cover spent fuel. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Methods and procedures used in different countries are described in this report. Refs, figs, tabs.

  1. DWARF, 1-D Few-Group Neutron Diffusion with Thermal Feedback for Burnup and Xe Oscillation

    International Nuclear Information System (INIS)

    Anderson, E.C.; Putnam, G.E.

    1975-01-01

    1 - Description of problem or function: DWARF allows one-dimensional simulation of reactor burnup and xenon oscillation problems in slab, cylindrical, or spherical geometry using a few-group diffusion theory model. 2 - Method of solution: The few-group, neutron diffusion theory equations are reduced to a system of finite-difference equations that are solved for each group by the Gauss method at each time point. Fission neutron source iteration can be accelerated with Chebyshev extrapolation. A thermal feedback iterative loop is used to obtain consistent solutions for the distributions of reactor power, neutron flux, and fuel and coolant properties with the neutron group constants functions of the latter. Solutions for the new nuclide concentrations of a time-point are made with the flux assumed constant in the time interval. 3 - Restrictions on the complexity of the problem - Maxima of: 4 groups; 40 regions; 50 macroscopic materials (Only 10 are functions of the feedback variables); 50 nuclides per region; 250 mesh points

  2. BASHAN: A few-group three-dimensional diffusion code with burnup and fuel management features

    International Nuclear Information System (INIS)

    Pearce, D.F.

    1970-12-01

    The diffusion equation for a two or three-dimensional, two-group or multi-group downscatter problem is solved by conventional finite difference techniques. An x-y-z geometry is assumed with an 'in-channel' mesh point representation. Options are available which allow representation of a soluble poison dispersed throughout the reactor, and also absorber rods in specified channels. The power distribution and multiplication factor k eff are calculated and a point rating map is used to advance the irradiation at each mesh point by a specified time-step so that burnup is followed. Fuel changes may be made so that radial shuffling and axial shuffling fuel management schemes can be studies. The code has been written in FORTRAN S2 for an IBM 7030 (STRETCH) computer which, with a fast store of 80,000 locations, allows problems of up to 15,000 mesh points to be dealt with. Conversion to FORTRAN IV for IBM 360 has now been completed. (author)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

    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

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

    International Nuclear Information System (INIS)

    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

  5. Current applications of actinide-only burn-up credit within the Cogema group and R and D programme to take fission products into account

    International Nuclear Information System (INIS)

    Toubon, H.; Guillou, E.; Cousinou, P.; Barbry, F.; Grouiller, J.P.; Bignan, G.

    2001-01-01

    Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis

  6. Current applications of actinide-only burn-up credit within the Cogema group and R and D programme to take fission products into account

    Energy Technology Data Exchange (ETDEWEB)

    Toubon, H. [Cogema, 78 - Saint Quentin en Yvelines (France); Guillou, E. [Cogema Etablissement de la Hague, D/SQ/SMT, 50 - Beaumont Hague (France); Cousinou, P. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, 92 (France); Barbry, F. [CEA Valduc, Inst. de Protection et de Surete Nucleaire, 21 - Is sur Tille (France); Grouiller, J.P.; Bignan, G. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

    2001-07-01

    Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis.

  7. High Burnup Effects Program

    International Nuclear Information System (INIS)

    Barner, J.O.; Cunningham, M.E.; Freshley, M.D.; Lanning, D.D.

    1990-04-01

    This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water reactor (LWR) fuel irradiated to high burnup levels. The HBEP was organized into three tasks as follows: Task 1 -- high burnup effects evaluations; Task 2 -- fission gas sampling; and Task 3 -- parameter effects study. During the course of the HBEP, a program that extended over 10 years, 82 fuel rods from a variety of sources were characterized, irradiated, and then examined in detail after irradiation. The study of fission gas release at high burnup levels was the principal objective of the program and it may be concluded that no significant enhancement of fission gas release at high burnup levels was observed for the examined rods. The rim effect, an as yet unquantified contributor to athermal fission gas release, was concluded to be the one truly high-burnup effect. Though burnup enhancement of fission gas release was observed to be low, a full understanding of the rim region and rim effect has not yet emerged and this may be a potential area of further research. 25 refs., 23 figs., 4 tabs

  8. Appropriate burnup measurements for transportation burnup credit

    International Nuclear Information System (INIS)

    Lancaster, D.; Fuentes, E.

    1997-01-01

    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

  9. Impacts of burnup-dependent swelling of metallic fuel on the performance of a compact breed-and-burn fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Heo, Woong; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-08-01

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

  11. Calculation of burnup and power dependence on fission gas released from PWR type reactor fuel element

    International Nuclear Information System (INIS)

    Edy-Sulistyono

    1996-01-01

    Burn up dependence of fission gas released and variation power analysis have been conducted using FEMXI-IV computer code program for Pressure Water Reactor Fuel During steady-state condition. The analysis result shows that the fission gas release is sensitive to the fuel temperature, the increasing of burn up and power in the fuel element under irradiation experiment

  12. Research on burnup physics

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1974-07-01

    One of the major problems in burnup studies is the reasonably fast and accurate calculation of the space-and-energy dependent neutron flux and reaction rates for realistic power reactor fuel geometries and compositions, and its optimal integration in the global reactor calculations. The scope of the present research was to develop improved methods trying to satisfy the above requirements. In the epithermal region, simple and efficient approximation is proposed which allows the analytical solution for the space dependence of the spherical harmonics flux moments, and hence the derivation of the recurrence relations between he flux moments at successive lethargy pivotal points. A new matrix formalism to invert the coefficient matrix of band structure resulted in a reduce computer time and memory demands. The research on epithermal region is finalized in computing programme SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, and the related integral quantities as reaction rates and resonance integrals. For partial verification of the above methods a Monte Carlo procedure was developed. Using point-wise representation of variables, a flexible and fast convergent integral transport method SEPT i developed. Expanding the neutron source and flux in finite series of arbitrary polynomials, the space-and-energy dependent integral transport equation is transformed into a general linear algebraic form, which is solved numerically. A simple and efficient procedure for deriving multipoint equations and constructing matrix is proposed and examined, and no unwanted oscillations were noticed. The energy point method was combined with the spherical harmonics method as well. A multi zone few-group program SPECTAR for global reactor calculations was developed. For testing, the flux distribution, neutron leakage and effective multiplication factor for the PWR reactor of the power station San Onofre were calculated. In order to verify

  13. PLUTON: A Three-Group Model for the Radial Distribution of Plutonium, Burnup, and Power Profiles in Highly Irradiated LWR Fuel Rods

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Nakamura, Jinichi; Suzuki, Motoe

    2001-01-01

    A three-group model (PLUTON) is described, which predicts the power density distribution, plutonium buildup, and burnup profiles across the fuel pellet radius as a function of in-pile time and parameters characterizing the type of reactor system with respect to fuel temperature and changes of density during the irradiation period. The PLUTON model is a part of two fuel performance codes (ASFAD and FEMAXI-V), which provide all necessary input for this model, mainly local temperatures and fuel matrix density across the radius. Comparisons between measurements and predictions of the PLUTON model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnup between 21 000 and 64 000 MWd/t. It is shown that the PLUTON predictions are in good agreement with measurements as well as with predictions of the well-known TUBRNP model. The proposed model is flexibly applicable for all types of light water reactor (LWR) fuels, including mixed oxide, and for fuel tested in the Organization for Economic Corporation and Development's Halden heavy water reactor. The PLUTON three-group model is based on analytical (theoretical) consideration of neutron absorption in a resonant region of the fuel in its apparent form. It makes the model more flexible in comparison with the semi-empirical TUBRNP one-group model and allows the physically based model analysis of commercial LWR-type fuels at high burnup as well as analysis of experimental fuel rods tested in the Halden heavy water reactor, which is one of the main test reactors in the world. The differences in fuel behavior in the Halden reactor in terms of burnup distribution and plutonium buildup can be more clearly understood with the PLUTON model

  14. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

    Ahlf, J.

    1983-01-01

    The specialists' group for fuel elements of the Kerntechnische Gesellschaft e.V. held a meeting on ''Increased Burnup of Fuel Elements'' on 9th and 10th of November 1982 at the GKSS Research Center Geesthacht. Most papers dealt with the problems of burnup increase of fuel elements for light water reactors with respect to fuel manufacturing, power plant operation and reprocessing. Review papers were given on the burnup limits for high temperature gas cooled reactors and sodium fast breeder reactors. The meeting ended with a presentation of the technical equipment of the hot laboratory of the GKSS and the programs which are in progress there. (orig.) [de

  15. DELIGHT-B/REDEL, point reactivity burnup code for high-temperature gas-cooled reactor cells

    International Nuclear Information System (INIS)

    Shindo, Ryuiti; Watanabe, Takashi.

    1977-03-01

    Code DELIGHT-2 was previously developed to analyze cell burnup characteristics and to produce few-group constants for core burnup calculation in high-temperature gas-cooled reactors. In the code, burnup dependency of the burnable poison, boron-10, is considered with the homogeneous model of space. In actuality, however, the burnable poison is used as homogeneous rods or uniform rods of small granular poison and graphite, to control the reactivity and power distribution. Precise analysis of the burnup characteristics is thus difficult because of the heterogeneity due to the configuration of poison rods. In cell burnup calculation, the DELIGHT-B, which is a modification of DELIGHT-2, takes into consideration this heterogeneous effect. The auxiliary code REDEL, a reduction of DELIGHT-B, used in combination with 3 dimensional diffusion code CITATION, is for core burnup calculation with the macro-scopic cross section model. (auth.)

  16. High burnup issues and modelling strategies

    International Nuclear Information System (INIS)

    Dutta, B.K.

    2005-01-01

    The performance of high burnup fuel is affected by a number of phenomena, such as, conductivity degradation, modified radial flux profile, fission gas release from high burnup structures, PCMI, burnup dependent thermo-mechanical properties, etc. The modelling strategies of some of these phenomena are available in literature. These can be readily incorporated in a fuel modelling performance code. The computer code FAIR has been developed in BARC over the years to evaluate the fuel performance at extended burnup and modelling of the fuel rods for advanced fuel cycles. The present paper deals with the high burnup issues in the fuel pins, their modelling strategies and results of the case studies specifically involving high burnup fuel. (author)

  17. Burnup credit activities in the United States

    International Nuclear Information System (INIS)

    Lake, W.H.; Thomas, D.A.; Doering, T.W.

    2001-01-01

    This report covers progress in burnup credit activities that have occurred in the United States of America (USA) since the International Atomic Energy Agency's (IAEA's) Advisory Group Meeting (AGM) on Burnup Credit was convened in October 1997. The Proceeding of the AGM were issued in April 1998 (IAEA-TECDOC-1013, April 1998). The three applications of the use of burnup credit that are discussed in this report are spent fuel storage, spent fuel transportation, and spent fuel disposal. (author)

  18. Application of burnup credit concept to transport

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Nakagome, Yoshihiro.

    1994-01-01

    For the design and safety assessment of the casks for transporting spent fuel, the fuel contained in them has been assumed to be new fuel. The reason is, it was difficult to evaluate the variation of the reactivity of fuel, and the research on the affecting factors and the method of measuring burnup were not much advanced. Recently, high burnup fuel has been adopted, and initial degree of enrichment rose. The research has been advanced for pursuing the economy of the casks for spent fuel, and burnup credit has become applicable to their design and safety assessment. As the result, the containing capacity increases by about 20%. When burnup credit is considered, it is necessary to confirm accurately the burnup of spent fuel. The burnup dependence of the concentration of fissile substances and neutron emissivity, the coolant void dependence of the concentration of fissile substances, and the relation of neutron multiplication rate with initial degree of enrichment or burnup are discussed. The conceptual design of casks considering burnup credit and its assessment, the merit, problem and the countermeasures to it when burnup credit is introduced are described. (K.I.)

  19. Epistemic dependence in interdisciplinary groups

    DEFF Research Database (Denmark)

    Andersen, Hanne; Wagenknecht, Susann

    2013-01-01

    In interdisciplinary research scientists have to share and integrate knowledge between people and across disciplinary boundaries. An important issue for philosophy of science is to understand how scientists who work in these kinds of environments exchange knowledge and develop new concepts...... and theories across diverging fields. There is a substantial literature within social epistemology that discusses the social aspects of scientific knowledge, but so far few attempts have been made to apply these resources to the analysis of interdisciplinary science. Further, much of the existing work either...... ignores the issue of differences in background knowledge, or it focuses explicitly on conflicting background knowledge. In this paper we provide an analysis of the interplay between epistemic dependence between individual experts with different areas of expertise. We analyze the cooperative activity...

  20. Choosing the optimum burnup

    International Nuclear Information System (INIS)

    Geller, L.; Goldstein, L.; Franks, W.A.

    1986-01-01

    This paper reviews some of the considerations utilities must evaluate when going to higher discharge burnups. The advantages and disadvantages of higher discharge burnups are described, as well as a consistent approach for evaluating optimum discharge burnup and its comparison to current practice. When an analysis is performed over the life of the plant, the design of the terminal cycles has significant impact on the lifetime savings from higher burnups. Designs for high burnup cycles have a greater average inventory value in the core. As one goes to higher burnup, there is a greater likelihood of discarding a larger value in unused fuel unless the terminal cycles are designed carefully. This effect can be large enough in some cases to wipe out the lifetime cost savings relative to operating with a higher discharge burnup cycle

  1. PENBURN - A 3-D Zone-Based Depletion/Burnup Solver

    International Nuclear Information System (INIS)

    Manalo, Kevin; Plower, Thomas; Rowe, Mireille; Mock, Travis; Sjoden, Glenn E.

    2008-01-01

    PENBURN (Parallel Environment Burnup) is a general depletion/burnup solver which, when provided with zone-based reaction rates, computes time-dependent isotope concentrations for a set of actinides and fission products. Burnup analysis in PENBURN is performed with a direct Bateman-solver chain solution technique. Specifically, in tandem with PENBURN is the use of PENTRAN, a parallel multi-group anisotropic Sn code for 3-D Cartesian geometries. In PENBURN, the linear chain method is actively used to solve individual isotope chains which are then fully attributed by the burnup code to yield integrated isotope concentrations for each nuclide specified. Included with the discussion of code features, a single PWR fuel pin calculation with the burnup code is performed and detailed with a benchmark comparison to PIE (Post-Irradiation Examination) data within the SFCOMPO (Spent Fuel Composition / NEA) database, and also with burnup codes in SCALE5.1. Conclusions within the paper detail, in PENBURN, the accuracy of major actinides, flux profile behavior as a function of burnup, and criticality calculations for the PWR fuel pin model. (authors)

  2. Fuel burnup analysis for the Moroccan TRIGA research reactor

    International Nuclear Information System (INIS)

    El Bakkari, B.; El Bardouni, T.; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Zoubair, M.

    2013-01-01

    Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of

  3. Issues for effective implementation of burnup credit

    International Nuclear Information System (INIS)

    Parks, C.V.; Wagner, J.C.

    2001-01-01

    In the United States, burnup credit has been used in the criticality safety evaluation for storage pools at pressurized water reactors (PWRs) and considerable work has been performed to lay the foundation for use of burnup credit in dry storage and transport cask applications and permanent disposal applications. Many of the technical issues related to the basic physics phenomena and parameters of importance are similar in each of these applications. However, the nuclear fuel cycle in the United States has never been fully integrated and the implementation of burnup credit to each of these applications is dependent somewhat on the specific safety bases developed over the history of each operational area. This paper will briefly review the implementation status of burnup credit for each application area and explore some of the remaining issues associated with effective implementation of burnup credit. (author)

  4. Lattice cell burnup calculation

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1977-01-01

    Accurate burnup prediction is a key item for design and operation of a power reactor. It should supply information on isotopic changes at each point in the reactor core and the consequences of these changes on the reactivity, power distribution, kinetic characters, control rod patterns, fuel cycles and operating strategy. A basic stage in the burnup prediction is the lattice cell burnup calculation. This series of lectures attempts to give a review of the general principles and calculational methods developed and applied in this area of burnup physics

  5. Burnup credit feasibility for BWR spent fuel shipments

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1990-01-01

    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 of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. 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. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

  6. A Monte Carlo burnup code linking MCNP and REBUS

    International Nuclear Information System (INIS)

    Hanan, N. A.

    1998-01-01

    The REBUS-3 burnup code, used in the ANL RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult burnup analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented

  7. Configuration of LWR fuel enrichment or burnup yielding maximum power

    International Nuclear Information System (INIS)

    Bartosek, V.; Zalesky, K.

    1976-01-01

    An analysis is given of the spatial distribution of fuel burnup and enrichment in a light-water lattice of given dimensions with slightly enriched uranium, at which the maximum output is achieved. It is based on the spatial solution of neutron flux using a one-group diffusion model in which linear dependence may be expected of the fission cross section and the material buckling parameter on the fuel burnup and enrichment. Two problem constraints are considered, i.e., the neutron flux value and the specific output value. For the former the optimum core configuration remains qualitatively unchanged for any reflector thickness, for the latter the cases of a reactor with and without reflector must be distinguished. (Z.M.)

  8. Fission-product burnup chain model for research reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup; Lee, Jong Tai [Korea Atomic Energy Research Inst., Daeduk (Republic of Korea)

    1990-12-01

    A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the preudo-element and the preudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.(Author).

  9. Burnup credit in Spain

    International Nuclear Information System (INIS)

    Conde, J.M.; Recio, M.

    2001-01-01

    The status of development of burnup credit for criticality safety analyses in Spain is described in this paper. Ongoing activities in the country in this field, both national and international, are resumed. Burnup credit is currently being applied to wet storage of PWR fuel, and credit to integral burnable absorbers is given for BWR fuel storage. It is envisaged to apply burnup credit techniques to the new generation of transport casks now in the design phase. The analysis methodologies submitted for the analyses of PWR and BWR fuel wet storage are outlined. Analytical activities in the country are described, as well as international collaborations in this field. Perspectives for future research and development of new applications are finally resumed. (author)

  10. Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi; Kaneko, Toshiyuki.

    1996-12-01

    The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code 'MULTI-KENO' and the routine for the burnup calculation of the one dimensional burnup code 'UNITBURN'. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

  11. Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

    Energy Technology Data Exchange (ETDEWEB)

    Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Toshiyuki

    1996-12-01

    The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code `MULTI-KENO` and the routine for the burnup calculation of the one dimensional burnup code `UNITBURN`. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

  12. Triton burnup in JET

    International Nuclear Information System (INIS)

    Chipsham, E.; Jarvis, O.N.; Sadler, G.

    1989-01-01

    Triton burnup measurements have been made at JET using time-integrated copper activation and time-resolved silicon detector techniques. The results confirm the classical nature of both the confinement and the slowing down of the 1 MeV tritons in a plasma. (author) 8 refs., 3 figs

  13. CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback

    International Nuclear Information System (INIS)

    Ahnert, Carol; Aragones, Jose M.

    1983-01-01

    1 - Description of problem or function: CARMEN is a system of programs developed for the neutronic calculation of PWR cycles. It includes the whole chain of analysis from cell calculations to core calculations with burnup. The core calculations are based on diffusion theory with cross sections depending on the relevant space-dependent feedback effects which are present at each moment along the cycles. The diffusion calculations are in one, two or three dimensions and in two energy groups. The feedback effects which are treated locally are: burnup, water density, power density and fission products. In order to study in detail these parameters the core should be divided into as many zones as different cross section sets are expected to be required in order to reproduce reality correctly. A relevant difference in any feedback parameter between zones produces different cross section sets for the corresponding zones. CARMEN is also capable to perform the following calculations: - Multiplication factor by burnup step with fixed boron concentration - Buckling and control rod insertion - Buckling search by burnup step - Boron search by burnup step - Control rod insertion search by burnup step. 2 - Method of solution: The cell code (LEOPARD-TRACA) generates the fuel assembly cross sections versus burnup. This is the basic library to be used in the CARMEN code proper. With a planar distribution guess for power density, water density and fluxes, the macroscopic cross sections by zone are calculated by CARMEN, and then a diffusion calculation is done in the whole geometry. With the distribution of power density, heat accumulated in the coolant and the thermal and fast fluxes determined in the diffusion calculation, CARMEN calculates the values of the most relevant parameters that influence the macroscopic cross sections by zone: burnup, water density, effective fuel temperature and fission product concentrations. If these parameters by zone are different from the reference

  14. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

    Suyama, Kenya; Hirakawa, Naohiro; Iwasaki, Tomohiko.

    1997-11-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user's manual of SWAT. (author)

  15. Triton burnup in JET - profile effects

    International Nuclear Information System (INIS)

    Jarvis, O.N.; Conroy, S.W.; Marcus, F.B.; Sadler, G.J.; Belle, P. van

    1991-01-01

    Measurements of the 14 MeV neutron emission from triton burnup show that the 14 MeV emission profile shadows closely the 2,5 MeV profile but after a delay corresponding to the triton slowing down time. The slightly greater width of the 14 MeV neutron profile is a consequence of the finite Larmor radius of the tritons. It has not so far been possible to identify unambiguously any effects on the triton burnup that are attributable to sawtooth crashes. Finally, the time dependence of the triton profile indicates that the triton diffusion coefficient is very small ( 2 /s). (author) 4 refs., 3 figs

  16. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  17. On-line extraction of the variance caused by burn-up in in-core three-dimensional power distribution

    International Nuclear Information System (INIS)

    Wang Yaqi; Luo Zhengpei; Li Fu; Liu Wenfeng

    2001-01-01

    In most of PWRs, the ex-core ion-chambers are the sole real-time sensors to respond to in-core power and its axial offset. However, the calibration coefficient of the ion-chambers depends on the (3D) power distribution and varies with the burn-up. People expect to know the variance in distribution caused by burn-up directly from the signals of ion-chambers. This expectation is not realized as yet, because an ion-chamber almost only responds to its nearest fuel assemblies. The authors then developed a two-step method for burn-up characteristic extraction: the harmonics synthesis method and harmonics' burn-up grouping. Using the extracted burn-up characteristics, the relationship between the readings of the ex-core ion-chambers and the in-core 3D power distribution is set up. Through the simulation on the heating reactor, the method of burn-up characteristic extraction is verified under engineering conditions. It is possible to on-line extract the variance caused by burn-up in 3D power distribution

  18. A Monte Carlo burnup code linking MCNP and REBUS

    International Nuclear Information System (INIS)

    Hanan, N.A.; Olson, A.P.; Pond, R.B.; Matos, J.E.

    1998-01-01

    The REBUS-3 burnup code, used in the anl RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented. (author)

  19. COGEMA/TRANSNUCLEAIRE's experience with burnup credit

    International Nuclear Information System (INIS)

    Chanzy, Y.; Guillou, E.

    1998-01-01

    Facing a continuous increase in the fuel enrichments, COGEMA and TRANSNUCLEAIRE have implemented step by step a burnup credit programme to improve the capacity of their equipment without major physical modification. Many authorizations have been granted by the French competent authority in wet storage, reprocessing and transport since 1981. As concerns transport, numerous authorizations have been validated by foreign competent authorities. Up to now, those authorizations are restricted to PWR Fuel type assemblies made of enriched uranium. The characterization of the irradiated fuel and the reactivity of the systems are evaluated by calculations performed with well qualified French codes developed by the CEA (French Atomic Energy Commission): CESAR as a depletion code and APPOLO-MORET as a criticality code. The authorizations are based on the assurance that the burnup considered is met on the least irradiated part of the fuel assemblies. Besides, the most reactive configuration is calculated and the burnup credit is restricted to major actinides only. This conservative approach allows not to take credit for any axial profile. On the operational side, the procedures have been reevaluated to avoid misloadings and a burnup verification is made before transport, storage and reprocessing. Depending on the level of burnup credit, it consists of a qualitative (go/no-go) verification or of a quantitative measurement. Thus the use of burnup credit is now a common practice in France and Germany and new improvements are still in progress: extended qualifications of the codes are made to enable the use of six selected fission products in the criticality evaluations. (author)

  20. ABB high burnup fuel

    International Nuclear Information System (INIS)

    Andersson, S.; Helmersson, S.; Nilsson, S.; Jourdain, P.; Karlsson, L.; Limback, M.; Garde, A.M.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

  1. Development of methods for burn-up calculations for LWR's

    International Nuclear Information System (INIS)

    Jaschik, W.

    1978-01-01

    This method is based on all burn-up depending data, namely particle densities and neutron spectra, being available in a burn-up library. This one is created by means of a small number of cell burn-up calculations which can easily be carried out and in which the heterogeneous cell structure and self-shielding effects can explicitly be accounted for. Then the cluster burn-up is simulated by adequate correlation of the burn-up data. The advantage of this method is given by - an exact determination of the real spectrum distribution in the individual fuel element clusters; - an exact determination of the burn-up related spectrum variations for each fuel rod and for each burn-up value obtained; - accounting for heterogeneity of the fuel rod cells and the self-shielding in the fuel; high accuracy of the results of a comparably low effort and - simple handling by largely automating the process of computation. Programed realization was achieved by establishing the RSYST modules ABRAJA, MITHOM, and SIMABB and their implementation within the code system. (orig./HP) [de

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

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jungmin; Shi, Yue-Jiang; Chung, Kyoung-Jae, E-mail: jkjlsh1@snu.ac.k; Hwang, Y. S. [Department of Nuclear Engineering, Seoul National University, Seoul 151-744 (Korea, Republic of); Cheon, MunSeong; Rhee, T.; Kim, Junghee [National Fusion Research Institute, Daejeon 34133 (Korea, Republic of); Kim, Jun Young [Korea University of Science and Technology, Daejeon 34133 (Korea, Republic of); Isobe, M.; Ogawa, K. [National Institute for Fusion Science, Toki-shi (Japan); SOKENDAI (The Graduate University for Advanced Studies), Toki-shi (Japan)

    2016-11-15

    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 {sup 3}He 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.

  3. Influence of FIMA burnup on actinides concentrations in PWR reactors

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.

  4. Triton burnup in JET - profile effects

    Energy Technology Data Exchange (ETDEWEB)

    Jarvis, O.N.; Conroy, S.W.; Marcus, F.B.; Sadler, G.J.; Belle, P. van (Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking); Adams, J.M.; Watkins, N. (AEA Industrial Technology, Harwell Laboratory (United Kingdom))

    1991-01-01

    Measurements of the 14 MeV neutron emission from triton burnup show that the 14 MeV emission profile shadows closely the 2,5 MeV profile but after a delay corresponding to the triton slowing down time. The slightly greater width of the 14 MeV neutron profile is a consequence of the finite Larmor radius of the tritons. It has not so far been possible to identify unambiguously any effects on the triton burnup that are attributable to sawtooth crashes. Finally, the time dependence of the triton profile indicates that the triton diffusion coefficient is very small (<<0.1 m[sup 2]/s). (author) 4 refs., 3 figs.

  5. Modelling comonotonic group-life under dependent decrement causes

    OpenAIRE

    Wang, Dabuxilatu

    2011-01-01

    Comonotonicity had been a extreme case of dependency between random variables. This article consider an extension of single life model under multiple dependent decrement causes to the case of comonotonic group-life.

  6. TRIGA criticality experiment for testing burn-up calculations

    International Nuclear Information System (INIS)

    Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz

    1999-01-01

    A criticality experiment with partly burned TRIGA fuel is described. 20 wt % enriched standard TRIGA fuel elements initially containing 12 wt % U are used. Their average burn-up is 1.4 MWd. Fuel element burn-up is calculated in 2-D four group diffusion approximation using TRIGLAV code. The burn-up of several fuel elements is also measured by reactivity method. The excess reactivity of several critical and subcritical core configurations is measured. Two core configurations contain the same fuel elements in the same arrangement as were used in the fresh TRIGA fuel criticality experiment performed in 1991. The results of the experiment may be applied for testing the computer codes used for fuel burn-up calculations. (author)

  7. Impact of extended burnup on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-04-01

    The Advisory Group Meeting was held in Vienna from 2 to 5 December 1991, to review, analyse, and discuss the effects of burnup extension in both light and heavy water reactors on all aspects of the fuel cycle. Twenty experts from thirteen countries participated in this meeting. There was consensus that both economic and environmental benefits are driving forces toward the achievement of higher burnups and that the present trend of burnup extension may be expected to continue. The extended burnup has been considered for the three main stages of the fuel cycle: the front end, in-reactor issues and the back end. Thirteen papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  8. Optimum Discharge Burnup and Cycle Length for PWRs

    International Nuclear Information System (INIS)

    Secker, Jeffrey R.; Johansen, Baard J.; Stucker, David L.; Ozer, Odelli; Ivanov, Kostadin; Yilmaz, Serkan; Young, E.H.

    2005-01-01

    This paper discusses the results of a pressurized water reactor fuel management study determining the optimum discharge burnup and cycle length. A comprehensive study was performed considering 12-, 18-, and 24-month fuel cycles over a wide range of discharge burnups. A neutronic study was performed followed by an economic evaluation. The first phase of the study limited the fuel enrichments used in the study to 235 U consistent with constraints today. The second phase extended the range of discharge burnups for 18-month cycles by using fuel enriched in excess of 5 wt%. The neutronic study used state-of-the-art reactor physics methods to accurately determine enrichment requirements. Energy requirements were consistent with today's high capacity factors (>98%) and short (15-day) refueling outages. The economic evaluation method considers various component costs including uranium, conversion, enrichment, fabrication and spent-fuel storage costs as well as the effect of discounting of the revenue stream. The resulting fuel cycle costs as a function of cycle length and discharge burnup are presented and discussed. Fuel costs decline with increasing discharge burnup for all cycle lengths up to the maximum discharge burnup considered. The choice of optimum cycle length depends on assumptions for outage costs

  9. Simulation of integral local tests with high-burnup fuel

    International Nuclear Information System (INIS)

    Gyori, G.

    2011-01-01

    The behaviour of nuclear fuel under LOCA conditions may strongly depend on the burnup-dependent fuel characteristics, as it has been indicated by recent integral experiments. Fuel fragmentation and the associated fission gas release can influence the integral fuel behaviour, the rod rupture and the radiological release. The TRANSURANUS fuel performance code is a proper tool for the consistent simulation of burnup-dependent phenomena during normal operation and the thermo-mechanical behaviour of the fuel rod in a subsequent accident. The code has been extended with an empirical model for micro-cracking induced FGR and fuel fragmentation and verified against integral LOCA tests of international projects. (author)

  10. FURNACE-J, 2-D Diffusion Burnup for Fast Reactors from JAERI Fast-Set

    International Nuclear Information System (INIS)

    Ikawa, Koji

    1984-01-01

    1 - Nature of physical problem solved: FURNACEJ is a two-dimensional diffusion-burnup code for use in the detailed burnup analysis of fast reactors. The code is an extension code of the FURNACE. There exists no essential difference between FURNACE and FURNACEJ. However, the latter can deal with JAERI-Fast-Set as its cross section library, while the former is designed to use ABBN set. Additionally, in FURNACEJ, group-dependent and -independent transverse buckling of each region can be computed and punched on cards, if desired. This is prepared for users so as to use them as input data for detailed two-dimensional x-y calculations. 2 - Restrictions on the complexity of the problem: Only r-z geometry is available

  11. CHAR and BURNMAC - burnup modules of the AUS neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1986-03-01

    In the AUS neutronics code system, the burnup module CHAR solves the nuclide depletion equations by an analytic technique in a number of spatial zones. CHAR is usually used as one component of a lattice burnup calculation but contains features which also make it suitable for some global burnup calculations. BURNMAC is a simple accounting module based on the assumption that cross sections for a rector zone depend only on irradiation. BURNMAC is used as one component of a global calculation in which burnup is achieved by interpolation in the cross sections produced from a previous lattice calculation

  12. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

    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 andW 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)

  13. Improvements for Monte Carlo burnup calculation

    Energy Technology Data Exchange (ETDEWEB)

    Shenglong, Q.; Dong, Y.; Danrong, S.; Wei, L., E-mail: qiangshenglong@tsinghua.org.cn, E-mail: d.yao@npic.ac.cn, E-mail: songdr@npic.ac.cn, E-mail: luwei@npic.ac.cn [Nuclear Power Inst. of China, Cheng Du, Si Chuan (China)

    2015-07-01

    Monte Carlo burnup calculation is development trend of reactor physics, there would be a lot of work to be done for engineering applications. Based on Monte Carlo burnup code MOI, non-fuel burnup calculation methods and critical search suggestions will be mentioned in this paper. For non-fuel burnup, mixed burnup mode will improve the accuracy of burnup calculation and efficiency. For critical search of control rod position, a new method called ABN based on ABA which used by MC21 will be proposed for the first time in this paper. (author)

  14. Fundamental burn-up mode in a pebble-bed type reactor

    International Nuclear Information System (INIS)

    Chen, Xue-Nong; Kiefhaber, Edgar; Maschek, Werner

    2008-01-01

    This paper deals with a pebble-bed type reactor, in which the fuel is loaded from one side (top) and discharged from the other side (bottom). A boundary value problem of a single group diffusion equation coupled with simplified burn-up equations is studied, where the natural radioactive decay processes are neglected in the burn-up modelling. An asymptotic burning wave solution is found analytically in the one-dimensional case, which is called as fundamental burn-up mode. Among this solution family there are two particular cases, namely, a classic fundamental solution with a zero burn-up and a partial solitary burn-up wave solution with a highest burn-up. An example of Th-U conversion is considered and the solutions are presented in order to show the mechanism of the burning wave. (author)

  15. Robustness and Vulnerability of Networks with Dynamical Dependency Groups.

    Science.gov (United States)

    Bai, Ya-Nan; Huang, Ning; Wang, Lei; Wu, Zhi-Xi

    2016-11-28

    The dependency property and self-recovery of failure nodes both have great effects on the robustness of networks during the cascading process. Existing investigations focused mainly on the failure mechanism of static dependency groups without considering the time-dependency of interdependent nodes and the recovery mechanism in reality. In this study, we present an evolving network model consisting of failure mechanisms and a recovery mechanism to explore network robustness, where the dependency relations among nodes vary over time. Based on generating function techniques, we provide an analytical framework for random networks with arbitrary degree distribution. In particular, we theoretically find that an abrupt percolation transition exists corresponding to the dynamical dependency groups for a wide range of topologies after initial random removal. Moreover, when the abrupt transition point is above the failure threshold of dependency groups, the evolving network with the larger dependency groups is more vulnerable; when below it, the larger dependency groups make the network more robust. Numerical simulations employing the Erdős-Rényi network and Barabási-Albert scale free network are performed to validate our theoretical results.

  16. Technical description of the burn-up software system MOP

    International Nuclear Information System (INIS)

    Schutte, C.K.

    1991-05-01

    The burn-up software system MOP is a research tool primary intended to study the behaviour of fission products in any reactor composition. Input data are multi-group cross-sections and data concerning the nuclide chains. An option is available to calculate a fundamental mode neutron spectrum for the specified reactor composition. A separate program can test the consistency of the specified nuclide chains. Options are available to calculate time-dependent cross-sections of lumped fission products and to take account of the leakage of gaseous fission products from the reactor core. The system is written in FORTRAN77 for a CYBER computer, using the operating system NOS/BE. The report gives a detailed technical description of the applied algorithms and the flow and storage of data. Information is provided for adapting the system to other computer configurations. (author). 5 refs.; 11 figs

  17. Effect of Spatial-Dependent Utility on Social Group Domination

    Science.gov (United States)

    Rodriguez, Nathaniel; Meyertholen, Andrew

    2012-02-01

    The mathematical modeling of social group competition has garnered much attention. We consider a model originated by Abrams and Strogatz [Nature 424, 900 (2003)] that predicts the extinction of one of two social groups. This model assigns a utility to each social group, which is constant over the entire society. We find by allowing this utility to vary over a society, through the introduction of a network or spatial dependence, this model may result in the coexistence of the two social groups.

  18. Determination of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    Kristak, J.; Vobecky, M.

    1973-01-01

    Samples containing a known content of 235 U were irradiated with several different neutron doses and activities were determined of radionuclides including 125 Sb, 144 Ce, 134 Cs, 154 Eu, 103 Ru, 95 Zr. The values thus obtained were divided by the 137 Cs activity value. The resulting neutron dose-dependent value is plotted into a calibration graph. The degree of nuclear fuel burn-up is obtained from the graph using an experimentally determined ratio of the activities of the above radionuclides. (B.S.)

  19. Fuel cycle cost considerations of increased discharge burnups

    International Nuclear Information System (INIS)

    Scherpereel, L.R.; Frank, F.J.

    1982-01-01

    Evaluations are presented that indicate the attainment of increased discharge burnups in light water reactors will depend on economic factors particular to individual operators. In addition to pure resource conserving effects and assuming continued reliable fuel performance, a substantial economic incentive must exist to justify the longer operating times necessary to achieve higher burnups. Whether such incentive will exist or not will depend on relative price levels of all fuel cycle cost components, utility operating practices, and resolution of uncertainties associated with the back-end of the fuel cycle. It is concluded that implementation of increased burnups will continue at a graduated pace similar to past experience, rather than finding universal acceptance of particular increased levels at any particular time

  20. TRIGA fuel element burnup determination by measurement and calculation

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.; Persic, A.; Jeraj, R.

    2000-01-01

    To estimate the accuracy of the fuel element burnup calculation different factors influencing the calculation were studied. To cover different aspects of burnup calculations, two in-house developed computer codes were used in calculations. The first (TRIGAP) is based on a one-dimensional two-group diffusion approximation, and the second (TRIGLAV) is based on a two-dimensional four-group diffusion equation. Both codes use WIMSD program with different libraries forunit-cell cross section data calculation. The burnup accumulated during the operating history of the TRIGA reactor at Josef Stefan Institute was calculated for all fuel elements. Elements used in the core during this period were standard SS 8.5% fuel elements, standard SS 12% fuel elements and highly enriched FLIP fuel elements. During the considerable period of operational history, FLIP and standard fuel elements were used simultaneously in mixed cores. (authors)

  1. Use of burnup credit for transportation and storage

    International Nuclear Information System (INIS)

    Sanders, T.L.; Ewing, R.I.; Lake, W.H.

    1991-01-01

    Burnup credit is the application of the effects of fuel burnup to nuclear criticality design. When burnup credit is considered in the design of storage facilities and transportation casks for spent fuel, the objectives are to reduce the requirements for storage space and to increase the payload of casks with acceptable nuclear criticality safety margins. The spent-fuel carrying capacities of previous-generation transport casks have been limited primarily by requirements to remove heat and/or to provide shielding. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced significantly. Thus a considerable weight margin is available to the designer for increasing the payload capacity. One method to achieve an increase in capacity is to reduce fuel assembly spacing. The amount of reduction in assembly spacing is limited by criticality and fuel support structural concerns. The optimum fuel assembly spacing provides the maximum cask loading within a basket that has adequate criticality control and sufficient structural integrity for regulatory accident scenarios. The incorporation of burnup credit in cask designs could result in considerable benefits in the transport of spent fuel. The acceptance of burnup credit for the design of transport casks depends on the resolution of system safety issues and the uncertainties that affect the determination of criticality safety margins. The remainder of this report will examine these issues and the integrated approach under way to resolve them. 20 refs., 2 figs

  2. Disposal criticality analysis methodology's principal isotope burnup credit

    International Nuclear Information System (INIS)

    Doering, T.W.; Thomas, D.A.

    2001-01-01

    This paper presents the burnup credit aspects of the United States Department of Energy Yucca Mountain Project's methodology for performing criticality analyses for commercial light-water-reactor fuel. The disposal burnup credit methodology uses a 'principal isotope' model, which takes credit for the reduced reactivity associated with the build-up of the primary principal actinides and fission products in irradiated fuel. Burnup credit is important to the disposal criticality analysis methodology and to the design of commercial fuel waste packages. The burnup credit methodology developed for disposal of irradiated commercial nuclear fuel can also be applied to storage and transportation of irradiated commercial nuclear fuel. For all applications a series of loading curves are developed using a best estimate methodology and depending on the application, an additional administrative safety margin may be applied. The burnup credit methodology better represents the 'true' reactivity of the irradiated fuel configuration, and hence the real safety margin, than do evaluations using the 'fresh fuel' assumption. (author)

  3. High burnup MOX fuel assembly

    International Nuclear Information System (INIS)

    Blanpain, P.; Brunel, L.

    1999-01-01

    From the outset, the MOX product was required to have the same performance as UO 2 in terms of burnup and operational flexibility. In fact during the first years the UO 2 managements could not be applied to MOX. The changeover to an AFA 2G type fuel allowed an improvement in NPP operational flexibility. The move to the AFA 3G design fuel will enable an increase in the burnup of the MOX assemblies to the level of the UO 2 ones ('MOX Parity' project). But the FRAMATOME fuel development objective does not stop at the obtaining of parity between the current MOX and UO 2 products: this parity must remain guaranteed and the MOX managements must evolve in the same way as the UO 2 managements. The goal of the MOX product development programmes underway with COGEMA and the CEA is the demonstration over the next 10 years of a fuel capable of reaching burnups of 70 GWD/T. The research programmes focus on the fission gas release aspect, with three issues explored: optimization of pellet microstructures and validation in experimental reactor ; build-up of experience feedback from fission gas release at elevated burnups in commercial reactors, both for current and experimental products; adaptation and qualification of the design models and tools, over the ranges and for the products concerned. The product arising from these development programmes should be offered on the market around 2010. While meeting safety requirements, it will cater for the needs of the utilities in terms of product reliability, personnel dosimetry and kWh output costs (increase in burnup, NPP maneuverability and availability, minimization of process waste). (authors)

  4. Analysis on burn-up behaviors for accelerator-driven sub-critical facility

    International Nuclear Information System (INIS)

    Liu Guisheng; Zhao Zhixiang; Zhang Baocheng; Shen Qinbiao; Ding Dazhao

    2000-01-01

    An analysis is performed on burn-up behaviors for accelerator-driven sub-critical reactor by means of the code PASC-1 for neutronics calculation, the code CBURN for burn-up calculation and 44 group constants is processed by CENDL-2 and ENDF/B-6 using NJOY-91.91

  5. End effects in the criticality analysis of burnup credit casks

    International Nuclear Information System (INIS)

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

    1990-01-01

    A study to evaluate the effect of axially dependent burnup on k eff has been performed as part of an effort to qualify procedures to be used in establishing burnup credit in shipping cask design and certification. This study was performed using a generic 31-element modular cast-iron cask (wall thickness 33.1 cm) with a 1-cm-thick borated stainless-steel basket for reactivity control. Fuel isotopics used here are those of the 17 x 17 Westinghouse assemblies from the North Anna Unit 1 reactor. Virginia Power (VP) provided detailed spatial isotopics for the fuel assemblies in-core at beginning-of-cycle 5 (BOC-5) as generated from their PDQ analyses. Twenty-two axial planes were defined in the original VP data. The isotopics used in this study were for a 3.41 initial wt % 235 U and an average burnup of 31.5 GWd/MTU

  6. Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

    International Nuclear Information System (INIS)

    Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

    2016-01-01

    Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

  7. Determination of reactor fuel burnup using passive neutron assay

    International Nuclear Information System (INIS)

    Kodeli, I.; Trkov, A.; Najzer, M.; Ertek, C.

    1988-01-01

    Passive neutron assay (PNA) method was developed to verify the fissile inventory of the irradiated reactor fuels. The characteristics of the method were studied at 'Jozef Stefan' Institute. The dependence of neutron source in the fuel on burnup, cooling time, initial enrichment and specific power were investigated and the accuracy of the method, using available computer codes was estimated. (author)

  8. COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System

    International Nuclear Information System (INIS)

    Suyama, K.; Masukawa, F.; Ido, M.; Enomoto, M.; Takyu, S.; Hara, T.

    2002-01-01

    1 - Description of program or function: Burn-up calculation of nuclear fuel. 2 - Methods: Matrix exponential method, Bateman Equation. 3 - Restrictions on the complexity of the problem: a) One-grouped cross section library should be prepared for the fuel system to be analyzed using UNITBURN. However, UNITBURN is not available now for UNIX systems. b) Gamma ray spectrometry calculation will fail using the attached piflib routine. This problem has already been rectified in the internal version. 4 - Typical running time: Two minutes for standard burn-up calculation on Sun ULTRA 30. 5 - Unusual features - a) Selection of Matrix exponential method, or Bateman Equation. b) JDDL, a detailed decay chain data based on ENSDF. 6 - Related or auxiliary programs: UNITBURN: Burnup calculation code unit cell system

  9. Findings of an international study on burnup credit

    International Nuclear Information System (INIS)

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

    1996-01-01

    Findings from a four year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses are presented in this paper. Approximately 20 participants from 11 countries have provided results for most problems. Four detailed benchmark problems for Pressurized Water Reactor (PWR) fuel have been completed and are summarized in this paper. Preliminary results from current work addressing burnup credit for Boiling Water Reactor (BWR) fuel will also be discussed as well as planned activities for additional benchmarks including Mixed-Oxide (MOX) fuels, subcritical benchmarks, international databases, and other activities

  10. Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-01

    EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.

  11. End effect Keff bias curve for actinide-only burnup credit casks

    International Nuclear Information System (INIS)

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

    1997-01-01

    A conservative end effect k eff bias curve for actinide-only burnup credit for spent fuel casks is presented in this paper. The k 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 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 eff could eliminate the need for axially burnup dependent analyses. 7 refs., 1 fig

  12. Optimization of TRU burnup in modular helium reactor

    International Nuclear Information System (INIS)

    Yonghee, Kim; Venneri, F.

    2007-01-01

    An optimization study of a single-pass TRU (transuranic) deep-burn (DB) has been performed for a block-type MHR (Modular Helium Reactor) proposed by General Atomics. Assuming a future equilibrium scenario of advanced LWRs, a high-burnup TRU vector is considered: 50 GWD/MTU and 5-year cooling. For 3-D equilibrium cores, the performance analysis is done by using a continuous energy Monte Carlo depletion code MCCARD. The core optimization is performed from the viewpoints of the core configuration, fuel management, TRISO fuel specification, and neutron spectrum. With regard to core configuration, two annular cores are investigated in terms of the neutron economy. A conventional radial shuffling scheme of fuel blocks is compared with an axial block shuffling strategy in terms of the fuel burnup and core power distributions. The impact of the kernel size of TRISO fuel is evaluated and a diluted kernel, instead of a conventional concentrated kernel, is introduced to maximize the TRU burnup by reducing the self-shielding effects of TRISO fuels. A higher graphite density is evaluated in terms of the fuel burnup. In addition, it is shown that the core power distribution can be effectively controlled by zoning of the packing fraction of TRISO fuels. We also have shown that a long-cycle DB-MHR core can be designed by using a small batch size for fuel reloading, at the expense of a marginal decrease of the TRU discharge burnup. Depending on the fuel management scheme, fuel specifications, and core parameters, the TRU burnup in an optimized DB-MHR core is over 60% in a single-pass irradiation campaign. (authors)

  13. Microstructural change and its influence on fission gas release in high burnup UO 2 fuel

    Science.gov (United States)

    Une, K.; Nogita, K.; Kashibe, S.; Imamura, M.

    1992-06-01

    The microstructural change of UO 2 fuel pellets (burnup: 6-83 GWd/t), base irradiated under LWR conditions, has been studied by detailed postirradiation examinations. The lattice parameter near the fuel rim in the irradiated UO 2 increased with burnup and appeared to become constant beyond about 50 GWd/t. This lattice dilation was mainly due to the accumulation of radiation induced point defects. Moreover, the dislocation density in the UO 2 matrix developed progressively with burnup, and eventually the tangled dislocations organized many sub-grain boundaries in the highest burnup fuel of 83 GWd/t. This sub-grain structure induced by accumulated radiation damage was compatible in appearance with SEM fractography results which revealed sub-divided grains of sub-micron size in as-fabricated grains. The influence of burnup on 85Kr release from the UO 2 fuels has been examined by means of a postirradiation annealing technique. The higher fractional release of high burnup fuels was mainly due to the burnup dependence of the fractional burst release evolved on temperature ramp. The fractional burst release was represented in terms of the square root of burnup from 6 to 83 GWd/t.

  14. The Width of High Burnup Structure in LWR UO2 Fuel

    International Nuclear Information System (INIS)

    Koo, Yang-Hyun; Lee, Byung-Ho; Oh, Jae-Yong; Sohn, Dong-Seong

    2007-01-01

    The measured data available in the open literature on the width of high burnup structure (HBS) in LWR UO 2 fuel were analyzed in terms of pellet average burnup, enrichment, and grain size. Dependence of the HBS width on pellet average burnup was shown to be divided into three regions; while the HBS width is governed by accumulation of fission damage (i.e., burnup) for burnup below 60 GWd/tU, it seems to be restricted to some limiting value of around 1.5 mm for burnup above 75 GWd/tU due to high temperature which might have caused extensive annealing of irradiation damage. As for intermediate burnup between 60 and 75 GWd/tU, although temperature would not have been so high as to induce extensive annealing, the microstructural damage could have been partly annealed, resulting in the reduction of the HBS width. It was found that both enrichment and grain size also affects the HBS width. However, as long as the pellet average burnup is lower than about 75 GWd/tU, the effect does not appear to be significant for the enrichment and grain size that are typically used in current LWR fuel. (authors)

  15. Establishing a PWR burn-up library

    International Nuclear Information System (INIS)

    Lutz, D.C.

    1981-01-01

    Starting out from data file ENDF/B IV /1/, a cross-section library has been established for the calculation of operating conditions in pressurized water reactors of the type used in BIBLIS B. The library includes macroscopic, homogenized 2-group cross-sections for all types of fuel elements used in this reactor, including those equipped with boron glass rods. For their calculation the previous irradiation of the fuel has been taken into consideration by approximation. Information on fuel consumption from cell burn-up calculations has been stored in a separate data file. It was designed as a base for the determination of cross sections to be used in the calculation of the incident ''main-steam pipe fracture''. For this library the description of cross sections as a function of the moderator status chose the water densities at 300 0 C/155 bar, 190 0 C/140 bar and 100 0 C/100 bar as fixed values. The burn-up library has been tested by a three-dimensional calculation for the 1sup(st) cycle of the BIBLIS B-reactor using program QUABOX /2/. This showed variances with the anticipated course concerning critically, which can be explained almost quantitatively by known deficiencies of the ENDF/b-IV library. (orig.) [de

  16. Fission-gas release in fuel performing to extended burnups in Ontario Hydro nuclear generating stations

    International Nuclear Information System (INIS)

    Floyd, M.R.; Novak, J.; Truant, P.T.

    1992-06-01

    The average discharge burnup of CANDU fuel is about 200 MWh/kgU. A significant number of 37-element bundles have achieved burnups in excess of 400 MWh/kgU. Some of these bundles have experienced failures related to their extended operation. To date, hot-cell examinations have been performed on fuel elements from nine 37-element bundles irradiated in Bruce NGS-A that have burnups in the range of 300-800 MWh/kgU. 1 Most of these have declining power histories from peak powers of up to 59 kW/m. Fission-gas releases of up to 26% have been observed and exhibit a strong dependence on fuel power. This obscures any dependence on burnup. The extent of fission-gas release at extended burnups was not predicted by low-burnup code extrapolations. This is attributed primarily to a reduction in fuel thermal conductivity which results in elevated operating temperatures. Reduced conductivity is due, at least in part, to the buildup of fission products in the fuel matrix. Some evidence of hyperstoichiometry exists, although this needs to be further investigated along with any possible relation to CANLUB graphite coating behaviour and sheath oxidation. Residual tensile sheath strains of up to 2% have been observed and can be correlated with fuel power/fission-gas release. SCC 2 -related defects have been observed in the sheath and endcaps of elements from bundles experiencing declining power histories to burnups in excess of 500 MWh/kgU. This indicates that the current recommended burnup limit of 450 MWh/kgU is justified. SCC-related defects have also been observed in ramped bundles having burnups < 450 MWh/kgU. Hence, additional guidelines are in place for power ramping extended-burnup fuel

  17. Burn-up Credit Criticality Safety Benchmark-Phase II-E. Impact of Isotopic Inventory Changes due to Control Rod Insertions on Reactivity and the End Effect in PWR UO2 Fuel Assemblies

    International Nuclear Information System (INIS)

    Neuber, Jens Christian; Tippl, Wolfgang; Hemptinne, Gwendoline de; Maes, Philippe; Ranta-aho, Anssu; Peneliau, Yannick; Jutier, Ludyvine; Tardy, Marcel; Reiche, Ingo; Kroeger, Helge; Nakata, Tetsuo; Armishaw, Malcom; Miller, Thomas M.

    2015-01-01

    The report describes the final results of the Phase II-E Burn-up Credit Criticality Benchmark conducted by the Expert Group on Burn-up Credit Criticality Safety. The objective of Phase II of the Burn-up Credit Criticality Safety programme is to study the impact of axial burn-up profiles of PWR UO 2 spent fuel assemblies on the reactivity of PWR UO 2 spent fuel assembly configurations. The objective of the Phase II-E benchmark was to study the impact of changes on the spent nuclear fuel isotopic composition due to control rod insertion during depletion on the reactivity and the end effect of spent fuel assemblies with realistic axial burn-up profiles for different control rod insertion depths ranging from 0 cm (no insertion) to full insertion (i.e. to the case that the fuel assemblies were exposed to control rod insertion over their full active length). For this purpose two axial burn-up profiles have been extracted from an AREVA-NP-GmbH-owned 17x17-(24+1) PWR UO 2 spent fuel assembly burn-up profile database. One profile has an average burn-up of 30 MWd/kg U, the other profile is related to an average burn-up of 50 MWd/kg U. Two profiles with different average burn-up values were selected because the shape of the burn-up profile is affected by the average burn-up and the end effect depends on the average burn-up of the fuel. The Phase II-E benchmark exercise complements the Phase II-C and Phase II-D benchmark exercises. In Phase II-D different irradiation histories were analysed using different control rod insertion histories during depletion as well as irradiation histories without control rod insertion. But in all the histories analysed a uniform distribution of the burn-up and hence a uniform distribution of the isotopic composition were assumed; and in all the histories including any usage of control rods full insertion of the control rods was assumed. In Phase II-C the impact of the asymmetry of axial burn-up profiles on the reactivity and the end effect of

  18. Application of depletion perturbation theory to fuel cycle burnup analysis

    International Nuclear Information System (INIS)

    White, J.R.

    1979-01-01

    Over the past several years static perturbation theory methods have been increasingly used for reactor analysis in lieu of more detailed and costly direct computations. Recently, perturbation methods incorporating time dependence have also received attention, and several authors have demonstrated their applicability to fuel burnup analysis. The objective of the work described here is to demonstrate that a time-dependent perturbation method can be easily and accurately applied to realistic depletion problems

  19. Phenomena and parameters important to burnup credit

    International Nuclear Information System (INIS)

    Parks, C.V.; Dehart, M.D.; Wagner, J.C.

    2001-01-01

    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 United States 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. (author)

  20. Burnup credit for storage and transportation casks

    International Nuclear Information System (INIS)

    Wells, A.H.

    1988-01-01

    The application of burnup credit to storage and transportation cask licensing results in a significant improvement in cask capacity and an associated reduction of the cost per kilogram of uranium in the cask contents. The issues for licensing with burnup credit deal primarily with the treatment of fission product poisons and methods of verification of burnup during cask operations. Other issues include benchmarking of cross-section sets and codes and the effect of spatial variation of burnup within an assembly. The licensing of burnup credit for casks will be complex, although the criticality calculations are not themselves difficult. Attention should be directed to the use of fission product poisons and the uncertainties that they introduce. Verification of burnup by measurements will remove some of the concerns for criticality safety. Calculations for burnup credit casks should consider rod-to-rod and axial variations of burnup, as well as variability of burnable poisons it they are used in the assembly. In spite of the complexity of cask burnup credit licensing issues, these issues appear to be resolvable within the current state of the art of criticality safety

  1. HAMCIND, Cell Burnup with Fission Products Poisoning

    International Nuclear Information System (INIS)

    Abe, Alfredo Y.; Dos Santos, Adimir

    2002-01-01

    1 - Description of program or function: HAMCIND is a cell burnup code based in a coupling between HAMMER-TECHNION and CINDER. The fission product poisoning is taken into account in an explicit fashion. 2 - Method of solution: The nonlinear coupled set of equations for the neutron transport and nuclide transmutation equations and nuclide transmutation equations in a unit cell is solved by HAMCIND in a quasi-static approach. The spectral transport equation is solved by HAMMER-TECHNION at the beginning of each time-step while the nuclide transmutation equations are solved by CINDER for every time-step. The HAMMER-TECHNION spectral calculations are performed taking into account the fission product contribution to the macroscopic cross sections (fast and thermal), in the inelastic scattering matrix and even in the thermal scattering matrices. 3 - Restrictions on the complexity of the problem: Restrictions and/or limitations for HAMCIND depend upon the local operating system

  2. Development of burnup methods and capabilities in Monte Carlo code RMC

    International Nuclear Information System (INIS)

    She, Ding; Liu, Yuxuan; Wang, Kan; Yu, Ganglin; Forget, Benoit; Romano, Paul K.; Smith, Kord

    2013-01-01

    Highlights: ► The RMC code has been developed aiming at large-scale burnup calculations. ► Matrix exponential methods are employed to solve the depletion equations. ► The Energy-Bin method reduces the time expense of treating ACE libraries. ► The Cell-Mapping method is efficient to handle massive amounts of tally cells. ► Parallelized depletion is necessary for massive amounts of burnup regions. -- Abstract: The Monte Carlo burnup calculation has always been a challenging problem because of its large time consumption when applied to full-scale assembly or core calculations, and thus its application in routine analysis is limited. Most existing MC burnup codes are usually external wrappers between a MC code, e.g. MCNP, and a depletion code, e.g. ORIGEN. The code RMC is a newly developed MC code with an embedded depletion module aimed at performing burnup calculations of large-scale problems with high efficiency. Several measures have been taken to strengthen the burnup capabilities of RMC. Firstly, an accurate and efficient depletion module called DEPTH has been developed and built in, which employs the rational approximation and polynomial approximation methods. Secondly, the Energy-Bin method and the Cell-Mapping method are implemented to speed up the transport calculations with large numbers of nuclides and tally cells. Thirdly, the batch tally method and the parallelized depletion module have been utilized to better handle cases with massive amounts of burnup regions in parallel calculations. Burnup cases including a PWR pin and a 5 × 5 assembly group are calculated, thereby demonstrating the burnup capabilities of the RMC code. In addition, the computational time and memory requirements of RMC are compared with other MC burnup codes.

  3. How groups contest depends on group power and the likelihood that power determines victory and defeat

    NARCIS (Netherlands)

    Kamans, Elanor; Otten, Sabine; Gordijn, Ernestine H.; Spears, Russell

    2010-01-01

    The aim of the current study was to show that the type of conflict behavior (constructive vs. unconstructive) groups use in conflicts depends on their power position as well as the likelihood that power determines victory and defeat. In an alleged online debate, we created a conflict between two

  4. The time-dependent density matrix renormalisation group method

    Science.gov (United States)

    Ma, Haibo; Luo, Zhen; Yao, Yao

    2018-04-01

    Substantial progress of the time-dependent density matrix renormalisation group (t-DMRG) method in the recent 15 years is reviewed in this paper. By integrating the time evolution with the sweep procedures in density matrix renormalisation group (DMRG), t-DMRG provides an efficient tool for real-time simulations of the quantum dynamics for one-dimensional (1D) or quasi-1D strongly correlated systems with a large number of degrees of freedom. In the illustrative applications, the t-DMRG approach is applied to investigate the nonadiabatic processes in realistic chemical systems, including exciton dissociation and triplet fission in polymers and molecular aggregates as well as internal conversion in pyrazine molecule.

  5. Enlarged Halden programme group meeting on high burn-up fuel performance, safety and reliability and degradation of in-core materials and water chemistry effects and man-machine systems research. Volume II

    International Nuclear Information System (INIS)

    1999-01-01

    Academy of Sciences, KFKI Atomic Energy Research Institute, the N.V. KEMA, the Netherlands, the Russian Research Centre 'Kurchatov Institute', the Slovakian VUJE - Nuclear Power Plant Research Institute, and from USA: the ABB Combustion Engineering Inc., the Electric Power Research Institute (EPRI), and the General Electric Co. The right to utilise information originating from the research work of the Halden Project is limited to persons and undertakings specifically given this right by one of these Project member organisations. The activities in the area of fuel and materials performance are based on extensive in-reactor measurements. The programmes are expanding in the areas of fuel performance at extended burn-ups, waterside corrosion and material testing in general. Development of in-core instruments is an important activity in support of the experimental programmes. The research programme at the Halden Project addresses the research needs of the nuclear industry in connection with introduction of digital I and C systems in NPPs. The programme provides information supporting design and licensing of upgraded, computer-based control room systems, and demonstrates the benefits of such systems through validation experiments in Halden's experimental research facility, HAMMLAB and pilot installations in NPPs. The Enlarged Halden Programme Group Meeting at Loen, Norway, was arranged to provide an opportunity to present results of work carried out at Halden and within participating organisations, and to encourage comments and impulses related to future Halden Project work. This HPR-351 relates to the fuel and materials part of the meeting and is divided in two volumes, HPR-351 Volume I and HPR-351 Volume II. The corresponding collection of papers in the man-machine area are given in one volume, HPR-352 Volume I. The overall programme of the Loen Enlarged Meeting covering the Fuel and Materials Research is given in the following pages. The papers with denomination HWR have

  6. Burnup determination of mass spectrometry for nuclear fuels

    International Nuclear Information System (INIS)

    Zhang Chunhua.

    1987-01-01

    The various methods currently being used in burnup determination of nuclear fuels are studied and reviewed. The mass spectrometry method of destructive testing is discussed emphatically. The burnup determination of mass spectrometry includes heavy isotopic abundance ratio method and isotope dilution mass spectrometry used as burnup indicator for the fission products. The former is applied to high burnup level, but the later to various burnup level. According to experiences, some problems which should be noticed in burnup determination of mass spectrometry are presented

  7. Experimental programmes related to high burnup fuel

    International Nuclear Information System (INIS)

    Vasudeva Rao, P.R.; Vidhya, R.; Ananthasivan, K.; Srinivasan, T.G.; Nagarajan, K.

    2002-01-01

    The experimental programmes undertaken at IGCAR with regard to high burn-up fuels fall under the following categories: a) studies on fuel behaviour, b) development of extractants for aqueous reprocessing and c) development of non-aqueous reprocessing techniques. An experimental programme to measure the carbon potential in U/Pu-FP-C systems by methane-hydrogen gas equilibration technique has been initiated at IGCAR in order to understand the evolution of fuel and fission product phases in carbide fuel at high burn-up. The carbon potentials in U-Mo-C system have been measured by this technique. The free energies and enthalpies of formation of LaC 2 , NdC 2 and SmC 2 have been measured by measuring the vapor pressures of CO over the region Ln 2 O 3 -LnC 2 -C during the carbothermic reduction of Ln 2 O 3 by C. The decontamination from fission products achieved in fuel reprocessing depends strongly on the actinide loading of the extractant phase. Tri-n-butyl phosphate (TBP), presently used as the extractant, does not allow high loadings due to its propensity for third phase formation in the extraction of Pu(IV). A detailed study of the allowable Pu loadings in TBP and other extractants has been undertaken in IGCAR, the results of which are presented in this paper. The paper also describes the status of our programme to develop a non-aqueous route for the reprocessing of fast reactor fuels. (author)

  8. Comparison of analysis methods for burnup credit applications

    International Nuclear Information System (INIS)

    Sanders, T.L.; Brady, M.C.; Renier, J.P.; Parks, C.V.

    1989-01-01

    The current approach used for the development and certification of spent fuel storage and transport casks requires an assumption of fresh fuel isotopics in the criticality safety analysis. However, it has been shown that there is a considerable reactivity reduction when the isotopics representative of the depleted (or burned) fuel are used in a criticality analysis. Thus, by taking credit for the burned state of the fuel (i.e., burnup credit), a cask designer could achieve a significant increase in payload. Accurate prediction of k eff for spent fuel arrays depends both on the criticality safety analysis and the prediction of the spent fuel isotopics via a depletion analysis. Spent fuel isotopics can be obtained from detailed multidimensional reactor analyses, e.g. the code PDQ, or from point reactor burnup models. These reactor calculations will help verify the adequacy of the isotopics and determine Δk eff biases for various analysis assumptions (with and without fission products, actinide absorbers, burnable poison rods, etc.). New software developed to interface PDQ multidimensional isotopics with KENO V.a reactor and cask models is described. Analyses similar to those performed for the reactor cases are carried out with a representative burnup credit cask model using the North Anna fuel. This paper presents the analysis methodology that has been developed for evaluating the physics issues associated with burnup credit. It is applicable in the validation and characterization of fuel isotopics as well as in determining the influence of various analysis assumptions in terms of δk eff . The methodology is used in the calculation of reactor restart criticals and analysis of a typical burnup credit cask

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

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)

    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)

  10. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

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

    2000-12-01

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

  11. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-01

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

  12. Burnup credit implementation in WWER spent fuel management systems: Status and future aspects

    International Nuclear Information System (INIS)

    Manolova, M.

    1998-01-01

    This paper describes the motivation for possible burnup credit implementation in WWER spent fuel management systems in Bulgaria. The activities being done are described, namely: the development and verification of a 3D few-group diffusion burnup model; the application of the KORIGEN code for evaluation of WWER fuel nuclear inventory during reactor core lifetime and after spent fuel discharge; using the SCALE modular system (PC Version 4.1) for criticality safety analyses of spent fuel storage facilities. Future plans involving such important tasks as validation and verification of computer systems and libraries for WWER burnup credit analysis are shown. (author)

  13. Sophistication of burnup analysis system for fast reactor (2)

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

    2010-10-01

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  14. Burnup effect on nuclear fuel cycle cost using an equilibrium model

    International Nuclear Information System (INIS)

    Youn, S. R.; Kim, S. K.; Ko, W. I.

    2014-01-01

    The degree of fuel burnup is an important technical parameter to the nuclear fuel cycle, being sensitive and progressive to reduce the total volume of process flow materials and eventually cut the nuclear fuel cycle costs. This paper performed the sensitivity analysis of the total nuclear fuel cycle costs to changes in the technical parameter by varying the degree of burnups in each of the three nuclear fuel cycles using an equilibrium model. Important as burnup does, burnup effect was used among the cost drivers of fuel cycle, as the technical parameter. The fuel cycle options analyzed in this paper are three different fuel cycle options as follows: PWR-Once Through Cycle(PWR-OT), PWR-MOX Recycle, Pyro-SFR Recycle. These fuel cycles are most likely to be adopted in the foreseeable future. As a result of the sensitivity analysis on burnup effect of each three different nuclear fuel cycle costs, PWR-MOX turned out to be the most influenced by burnup changes. Next to PWR-MOX cycle, in the order of Pyro-SFR and PWR-OT cycle turned out to be influenced by the degree of burnup. In conclusion, the degree of burnup in the three nuclear fuel cycles can act as the controlling driver of nuclear fuel cycle costs due to a reduction in the volume of spent fuel leading better availability and capacity factors. However, the equilibrium model used in this paper has a limit that time-dependent material flow and cost calculation is impossible. Hence, comparative analysis of the results calculated by dynamic model hereafter and the calculation results using an equilibrium model should be proceed. Moving forward to the foreseeable future with increasing burnups, further studies regarding alternative material of high corrosion resistance fuel cladding for the overall

  15. Protocol group education for family caregivers of elderly dependents

    Directory of Open Access Journals (Sweden)

    Cristina Anguita Carpio

    2012-03-01

    Full Text Available Family is the main care source for the dependent person. The act of looking after somebody, involves the development of multiple tasks, apart from spending a lot of time. This implies a series of requirements that would be able to damage the family caregivers’ quality of life, and definitively, their health. Objetive: training for family caregivers to provide quality care, in order to succeed in this project, we establish three specific aims: improving the information and training, increase self-care abilities and focus on resources and support services for informal caregivers. Methods: We are going to implement an educative intervention in a group of 12 informal caregivers of people over 65 years, dependent on chronic diseases and develop home care. The program will be composed of 10 sessions, each one of two-hour-long. The first session will consist of an introduction and the last session will be reserved to solve doubts and to deal with the assessment of the program. Throughout the rest of sessions, contents about training, self, resources and assistance services for caregivers will be proposed. In order to evaluate the efficiency of the program, a multiple choice questionnaire will be taken both al the beginning and at the end of the different sessions. In order to evaluate the human resources and the applied methodology, another questionnaire will be passed.

  16. Review of high burn-up RIA and LOCA database and criteria

    International Nuclear Information System (INIS)

    Vitanza, C.; Hrehor, M.

    2006-01-01

    This document is intended to provide regulators, their technical support organizations and industry with a concise review of existing fuel experimental data at RIA and LOCA conditions and considerations on how these data affect fuel safety criteria at increasing burn-up. It mostly addresses experimental results relevant to BWR and PWR fuel and it encompasses several contributions from the various experts that participated in the CSNI SEGFSM activities. It also covers the information presented at the joint CSNI/CNRA Topical Discussion on high burn-up fuel issues that took place on this subject in December 2004. The report is organized in the following way: the CABRI RIA database (14 tests), the NSRR database (26 tests) and other databases, RIA failure thresholds, comparison of failure thresholds for the HZP case, LOCA database ductility tests and quench tests, LOCA safety limit, provisional burn-up dependent criterion for Zr-4. The conclusions are as follows. On RIA, there is a well-established testing method and a significant and relatively consistent database from NSRR and Cabri tests, especially on high burn-up Zr-2 and Zr-4 cladding. It is encouraging that several correlations have been proposed for the RIA fuel failure threshold. Their predictions are compared and discussed in this paper for a representative PWR case. On LOCA, there are two different test methods, one based on ductility determinations and the other based on 'integral' quench tests. The LOCA database at high burn-up is limited to both testing methods. Ductility tests carried out with pre-hydrided non-irradiated cladding show a pronounced hydrogen effect. Data for actual high burn-up specimens are being gathered in various laboratories and will form the basis for a burn-up dependent LOCA limit. A provisional burn-up dependent criterion is discussed in the paper

  17. Value of burnup credit beyond actinides

    International Nuclear Information System (INIS)

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

    1997-01-01

    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

  18. Burnup verification using the FORK measurement system

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1994-01-01

    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 measurement system, designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program, has been used to verify reactor site records for burnup and cooling time for many years. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. This report deals with the application of the FORK system to burnup credit operations based on measurements performed on spent fuel assemblies at the Oconee Nuclear Station of Duke Power Company

  19. MCB. A continuous energy Monte Carlo burnup simulation code

    International Nuclear Information System (INIS)

    Cetnar, J.; Wallenius, J.; Gudowski, W.

    1999-01-01

    A code for integrated simulation of neutrinos and burnup based upon continuous energy Monte Carlo techniques and transmutation trajectory analysis has been developed. Being especially well suited for studies of nuclear waste transmutation systems, the code is an extension of the well validated MCNP transport program of Los Alamos National Laboratory. Among the advantages of the code (named MCB) is a fully integrated data treatment combined with a time-stepping routine that automatically corrects for burnup dependent changes in reaction rates, neutron multiplication, material composition and self-shielding. Fission product yields are treated as continuous functions of incident neutron energy, using a non-equilibrium thermodynamical model of the fission process. In the present paper a brief description of the code and applied methods are given. (author)

  20. Nuclide Importance and the Steady-State Burnup Equation

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Nemoto, Atsushi

    2000-01-01

    Conventional methods for evaluating some characteristic values of nuclides relating to burnup in a given neutron spectrum are reviewed in a mathematically systematic way, and a new method based on the importance theory is proposed. In this method, these characteristic values of a nuclide are equivalent to the importances of the nuclide. By solving the equation adjoint to the steady-state burnup equation with a properly chosen source term, the importances for all nuclides are obtained simultaneously.The fission number importance, net neutron importance, fission neutron importance, and absorbed neutron importance are evaluated and discussed. The net neutron importance is a measure directly estimating neutron economy, and it can be evaluated simply by calculating the fission neutron importance minus the absorbed neutron importance, where only the absorbed neutron importance depends on the fission product. The fission neutron importance and absorbed neutron importance are analyzed separately, and detailed discussions of the fission product effects are given for the absorbed neutron importance

  1. MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors

    International Nuclear Information System (INIS)

    Camiciola, P.; Cundari, D.; Montagnini, B.

    1992-01-01

    1 - Description of program or function: The program solves the 1-D time-dependent one and two group coarse-mesh neutron diffusion equations, coupled with the equations for the delayed-neutron precursor, in plane geometry. 2 - Method of solution: The program is based on a simple coarse-mesh cubic approximation formula for the spatial behaviour of the flux inside each interval. An implicit scheme (the time-integrated method) is used for the advancement of the solution. The resulting (block three-diagonal) matrix is inverted at each time step by Thomas' method. 3 - Restrictions on the complexity of the problem: Number of coarse- mesh intervals LE 80; number of material regions LE 10; number of delayed-neutron precursor groups LE 10. Typical mesh sizes range from 5 cm to 20 cm; typical step length (non-prompt critical transients) ranges from 0.005 to 0.1 seconds

  2. Threshold burnup for recrystallization and model for rim porosity in the high burnup UO2 fuel

    International Nuclear Information System (INIS)

    Lee, Byung Ho; Koo, Yang Hyun; Sohn, Dong Seong

    1998-01-01

    Applicability of the threshold burnup for rim formation was investigated as a function of temperature by Rest's model. The threshold burnup was the lowest in the intermediate temperature region, while on the other temperature regions the threshold burnup is higher. The rim porosity was predicted by the van der Waals equation based of the rim pore radius of 0.75μm and the overpressurization model on rim pores. The calculated centerline temperature is in good agreement with the measured temperature. However, more efforts seem to be necessary for the mechanistic model of the rim effect including rim growth with the fuel burnup

  3. Comparison of scale/triton and helios burnup calculations for high burnup LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S.; Mispagel, T.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

    2009-07-01

    The presented analyses provide information about the suitability of the lattice burnup code HELIOS and the recently developed code SCALE/TRITON for the prediction of isotopic compositions of high burnup LWR fuel. The accurate prediction of the isotopic inventory of high burnt spent fuel is a prerequisite for safety analyses in and outside of the reactor core, safe loading of spent fuel into storage casks, design of next generation spent fuel casks and for any consideration of burnup credit. Depletion analyses are performed with both burnup codes for PWR and BWR fuel samples which were irradiated far beyond 50 GWd/t within the LWR-PROTEUS Phase II project. (orig.)

  4. Development of high burnup nuclear fuel technology

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

    1987-04-01

    The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

  5. Physical models for high burnup fuel

    International Nuclear Information System (INIS)

    Kanyukova, V.; Khoruzhii, O.; Likhanskii, V.; Solodovnikov, G.; Sorokin, A.

    2003-01-01

    In this paper some models of processes in high burnup fuel developed in Src of Russia Troitsk Institute for Innovation and Fusion Research are presented. The emphasis is on the description of the degradation of the fuel heat conductivity, radial profiles of the burnup and the plutonium accumulation, restructuring of the pellet rim, mechanical pellet-cladding interaction. The results demonstrate the possibility of rather accurate description of the behaviour of the fuel of high burnup on the base of simplified models in frame of the fuel performance code if the models are physically ground. The development of such models requires the performance of the detailed physical analysis to serve as a test for a correct choice of allowable simplifications. This approach was applied in the SRC of Russia TRINITI to develop a set of models for the WWER fuel resulting in high reliability of predictions in simulation of the high burnup fuel

  6. Burnup analysis of the power reactor, 2

    International Nuclear Information System (INIS)

    Ezure, Hideo

    1975-09-01

    In burnup analysis of JPDR-1 with FLARE, it was found to have problems. The program FLORA was developed for solution of the problems. By their bench mark tests FLORA was found to be useful for three-dimensional thermal-hydro-dynamic analysis of BWRs. It was applied to analysis of the burnup of JPDR-1. The input data and option of FLORA were corrected on referring to the results of gammer probe tests for JPDR-1. The void, source and burnup distributions were calculated each month during the operation. The burnup distribution in three assemblies revealed by a destructive test agrees better with that by FLORA than by FLARE. It was shown that the distortion of power distribution around the control rods by FLORA was smaller and closer to that by the gammer probe tests than by FLARE, and the connector of fuel assemblies and the plugs in the reflector had much influence on the power distribution. (auth.)

  7. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

    Suyama, Kenya; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu.

    1997-06-01

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)

  8. Burnup calculation code system COMRAD96

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu

    1997-06-01

    COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, `Cross Section Treatment`, `Generation and Depletion Calculation`, and `Post Process`. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the {gamma} Spectrum on a terminal. This report is the general description and user`s manual of COMRAD96. (author)

  9. 'CANDLE' burnup regime after LWR regime

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Nagata, Akito

    2008-01-01

    CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium. (author)

  10. Comparative study on plutonium and MA recycling in equilibrium burnup and standard burnup of PWR

    International Nuclear Information System (INIS)

    Waris, Abdul; Kurniadi, Rizal; Su'ud, Zaki; Permana, Sidik

    2005-01-01

    The equilibrium burnup model is a powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor. However, this method needs to be verified since the method is not a standard tool. The present study aimed to compare the characteristics of plutonium recycling and plutonium and minor actinides (MA) recycling in PWR with the equilibrium burnup and the standard burnup. In order to become more comprehensive study, an influence of moderator-to-fuel volume ratio (MFR) changes by changing the pin-pitch of fuel cell has been evaluated. The MFR ranges from 0.5 to 4.0. For the equilibrium burnup we used equilibrium cell-burnup code. We have employed 1368 nuclides in the equilibrium calculation with 129 of them are heavy metals (HMs). For standard burnup, SRAC2002 code has been utilized with 26 HMs and 66 fission products (FPs). The JENDL 3.2 library has been employed for both burnup schemes. The uranium, plutonium and MA vector, which resulted from the equilibrium burnup are directly used as fuel input composition for the standard burnup calculation. Both burnup results demonstrate that plutonium recycling and plutonium and MA recycling can be conducted safer in tight lattice core. They are also show the similar trend in neutron spectrum, which become harder with the increasing number of recycled heavy nuclides as well as the decreasing of the MFR values. However, there are some discrepancy on the effective multiplication factor and the conversion ratio, especially for the reactor core for MFR ≥ 2.0. (author)

  11. BEAVRS full core burnup calculation in hot full power condition by RMC code

    International Nuclear Information System (INIS)

    Liu, Shichang; Liang, Jingang; Wu, Qu; Guo, JuanJuan; Huang, Shanfang; Tang, Xiao; Li, Zeguang; Wang, Kan

    2017-01-01

    Highlights: • TMS and thermal scattering interpolation were developed to treat cross sections OTF. • Hybrid coupling system was developed for HFP burnup calculation of BEAVRS benchmark. • Domain decomposition was applied to handle memory problem of full core burnup. • Critical boron concentration with burnup by RMC agrees with the benchmark results. • RMC is capable of multi-physics coupling for simulations of nuclear reactors in HFP. - Abstract: Monte Carlo method can provide high fidelity neutronics analysis of different types of nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with multi-physics interacting and coupling. MC codes can couple with depletion solver and thermal-hydraulics (T/H) codes simultaneously for the “transport-burnup-thermal-hydraulics” coupling calculations. MIT BEAVRS is a typical “transport-burnup-thermal-hydraulics” coupling benchmark. In this paper, RMC was coupled with sub-channel code COBRA, equipped with on-the-fly temperature-dependent cross section treatment and large-scale detailed burnup calculation based on domain decomposition. Then RMC was applied to the full core burnup calculations of BEAVRS benchmark in hot full power (HFP) condition. The numerical tests show that domain decomposition method can achieve the consistent results compared with original version of RMC while enlarging the computational burnup regions. The results of HFP by RMC agree well with the reference values of BEAVRS benchmark and also agree well with those of MC21. This work proves the feasibility and accuracy of RMC in multi-physics coupling and lifecycle simulations of nuclear reactors.

  12. Burnup calculations for cadmium. A case study for HFR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Sciolla, C.M

    2000-09-11

    This report describes the pre-design burnup calculations performed for a cadmium shielded high fluence irradiation experiment in the HFR. The very high absorption cross section in cadmium causes problems in the calculations for two different reasons. Firstly, because of the large reaction rates the assumption that the flux and the cross sections remain piecewise constant is no longer true. Therefore the correct solution can only be obtained when using extremely small time steps which leads to excessive computing times. Secondly, the self-shielding in the cadmium becomes complete (black absorber) causing the depletion to progress in a shell-wise manner. As a consequence the depletion evolves nearly linear instead of exponential with time. Because of this the depletion codes are used in a regime for which these have not been designed leading to a systematic error. The analysis shows however that a good estimate for the burnup time can be obtained by extrapolation from calculations with practically sized time steps and a correction is derived to compensate the systematic error. The calculations were done using the OCTOPUS burnup code system, including the 3-D Monte-Carlo spectrum code MCNP-4B and the depletion code FISPACT-4.2. Verifications were performed with the WIMS code system. The first part of the report describes the study of the cadmium burnup calculations for a shielded steel sample with the emphasis on analyzing the requirements for obtaining the correct solution. The second part describes the time-dependent power production calculations with the steel replaced by lithium containing ceramic material such as to be used in the 'High Fluence Irradiation of Ceramics for Fusion' (HICU) experiment. 12 refs.

  13. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2005-02-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For

  14. OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN

    International Nuclear Information System (INIS)

    Hesse, Ulrich; Sieberer, Johann

    2006-01-01

    1 - Description of program or function: In OREST, the 1-dimensional lattice code HAMMER and the isotope generation and depletion code ORIGEN are directly coupled for burnup simulation in light-water reactor fuels (GRS recommended). Additionally heavy water and graphite moderated systems can be calculated. New version differs from the previous version in the following features: An 84-group-library LIB84 for up to 200 isotopes is used to update the 3-group -POISON-XS. LIB84 uses the same energy boundaries as THERMOS and HAMLET in . In this way, high flexibility is achieved in very different reactor models. The coupling factor between THERMOS and HAMLET is now directly transferred from HAMMER to THERES and omits the equation 4 (see page 6 of the manual). Sandwich-reactor fuel reactivity and burnup calculations can be started with NGEOM = 1. Thorium graphite reactivity and burnup calculations can be started with NLIBE = 1. High enriched U-235 heavy water moderated reactivity and burnup calculations can be started. HAMLET libraries in for U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-242, Am-241, Am-243 and Zirconium are updated using resonance parameters. NEA-1324/04: A new version of the module hamme97.f has replaced the old one. 2 - Method of solution: For the user-defined irradiation history, an input data processor generates program loops over small burnup steps for the main codes HAMMER and ORIGEN. The user defined assembly description is transformed to an equivalent HAMMER fuel cell. HAMMER solves the integral neutron transport equation in a four-region cylindrical or sandwiched model with reflecting boundaries and runs with fuel power calculated rod temperatures. ORIGEN runs with HAMMER-calculated cross sections and neutron spectra and calculates isotope concentrations during burnup by solving the buildup-, depletion- and decay-chain equations. An output data processor samples the outputs of the program modules and generates tabular works for the

  15. Application of burnup credit for PWR spent fuel storage pool

    International Nuclear Information System (INIS)

    Shin, Hee Sung; Ro, Seung-Gy; Bae, Kang Mok; Kim, Ik Soo; Shin, Young Joon

    1999-01-01

    A study on the application of burnup credit for a PWR spent fuel storage pool has been investigated using a computer code system such as CSAS6 module of SCALE 4.3 in association with 44-group SCALE cross-section library. The calculation bias of the code system at a 95% probability with a 95% confidence level seems to be 0.00951 by benchmarking the system for forty six experimental data. With the aid of this computer code system, criticality analysis has been performed for the PWR spent fuel storage pool. Uncertainties due to postulated abnormal and accidental conditions, and manufacturing tolerance such as stainless steel thickness of storage rack, fuel enrichment, fuel density and box size have statistically been combined and resulted in 0.00674. Also, isotopic correction factor which was based on the calculated and measured concentration of 43 isotopes for both selected actinides and fission products important in burnup credit application has been taken into account in the criticality analysis. It is revealed that the minimum burnup with the corrected isotopic concentrations as required for the safe storage is 5,730 MWd/tU in enriched fuel of 5.0 wt%. (author)

  16. TEMPS, 1-Group Time-Dependent Pulsed Source Neutron Transport

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    1988-01-01

    1 - Description of program or function: TEMPS numerically determines the scalar flux as given by the one-group neutron transport equation with a pulsed source in an infinite medium. Standard plane, point, and line sources are considered as well as a volume source in the negative half-space in plane geometry. The angular distribution of emitted neutrons can either be isotropic or mono-directional (beam) in plane geometry and isotropic in spherical and cylindrical geometry. A general anisotropic scattering Kernel represented in terms of Legendre polynomials can be accommodated with a time- dependent number of secondaries given by c(t)=c 0 (t/t 0 ) β , where β is greater than -1 and less than infinity. TEMPS is designed to provide the flux to a high degree of accuracy (4-5 digits) for use as a benchmark to which results from other numerical solutions or approximations can be compared. 2 - Method of solution: A semi-analytic Method of solution is followed. The main feature of this approach is that no discretization of the transport or scattering operators is employed. The numerical solution involves the evaluation of an analytical representation of the solution by standard numerical techniques. The transport equation is first reformulated in terms of multiple collisions with the flux represented by an infinite series of collisional components. Each component is then represented by an orthogonal Legendre series expansion in the variable x/t where the distance x and time t are measured in terms of mean free path and mean free time, respectively. The moments in the Legendre reconstruction are found from an algebraic recursion relation obtained from Legendre expansion in the direction variable mu. The multiple collision series is evaluated first to a prescribed relative error determined by the number of digits desired in the scalar flux. If the Legendre series fails to converge in the plane or point source case, an accelerative transformation, based on removing the

  17. The US's group of seven: US oil import dependence grows

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    The latest available import data from the US Department of Energy show greater US dependence on imported crude oil, a two percentage point increase in 1992 above 1991. In fact, the share of imported crude oil of the sum of the US crude production and imports has grown from less than one-third in 1986 to nearly 46% in 1992. A closer look at the data reveals that US crude oil dependence is become increasingly centered on a small number of countries. In 1992, more than 85% of crude oil imports originated in seven countries: Saudia Arabia, Nigeria, Venezuela, Canada, Mexico, Angola, and the United Kingdom. Many people worry that US import dependence is at dangerous levels. However, the recent trend of joint ventures between US companies and oil producers should allay those fears by providing oil producers an incentive for stable oil supply in the US

  18. Investigation of the burn-up behavior of boron poison rods, placed in a fuel assembly of a pressurized water reactor

    International Nuclear Information System (INIS)

    Arnold, C.; Lutz, D.C.

    1979-09-01

    The excess reactivity of a pressurized water reactor is compensated by boron, disolved in the moderator. In addition during the first cycle boron poison rods are placed in fuel assemblies without control rods. The burn-up behavior of a poison rod in a Biblis B fuel assembly is analysed in the present paper. Multigroup spectrum calculations were performed. The influence of critical boron concentration depending from burn-up, the changes of fuel concentration and the concentration of burnable poison were taken into consideration. Furthermore the built-up of rapidly saturating fisson products 135 Xe and 149 Sm was considered. The interaction of these effects are discussed. Spatial influences are emphasized most. Finally two group cross sections were calculated. The results are compared with calculations for a fuel assembly of the same type without burnable poison rods. (orig.) [de

  19. OECD/NEA burnup credit criticality benchmarks phase IIIB: Burnup calculations of BWR fuel assemblies for storage and transport

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

    2002-02-01

    The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of ±10% relative to the average, although some results, esp. 155 Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k ∞ also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

  20. OECD/NEA burnup credit criticality benchmarks phase IIIB. Burnup calculations of BWR fuel assemblies for storage and transport

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-02-01

    The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

  1. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2004-02-01

    To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this

  2. Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

    International Nuclear Information System (INIS)

    Belo, Thiago F.; Fiel, Joao Claudio B.

    2015-01-01

    Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)

  3. Parameterized representation of macroscopic cross section in the PWR fuel element considering burn-up cycles

    Energy Technology Data Exchange (ETDEWEB)

    Belo, Thiago F.; Fiel, Joao Claudio B., E-mail: thiagofbelo@hotmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)

  4. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  5. Nuclear fuels with high burnup: safety requirements

    International Nuclear Information System (INIS)

    Phuc Tran Dai

    2016-01-01

    Vietnam authorities foresees to build 3 reactors from Russian design (VVER AES 2006) by 2030. In order to prepare the preliminary report on safety analysis the Vietnamese Agency for Radioprotection and Safety has launched an investigation on the behaviour of nuclear fuels at high burnups (up to 60 GWj/tU) that will be those of the new plants. This study deals mainly with the behaviour of the fuel assemblies in case of loss of coolant (LOCA). It appears that for an average burnup of 50 GWj/tU and for the advanced design of the fuel assembly (cladding and materials) safety requirements are fulfilled. For an average burnup of 60 GWj/tU, a list of issues remains to be assessed, among which the impact of clad bursting or the hydrogen embrittlement of the advanced zirconium alloys. (A.C.)

  6. Burnup credit in a dry storage module

    International Nuclear Information System (INIS)

    Thornton, J.R.

    1989-01-01

    Comparison of spent fuel storage expansion options available to Oconee Nuclear Station revealed that dry storage could be economically competitive with transshipment and rod consolidation. Economic competitiveness, however, mandated large unit capacity while existing cask handling facilities at Oconee severely limited size and weight. The dry storage concept determined to best satisfy these conflicting criteria is a 24 pressurized water reactor (PWR) fuel assembly capacity NUTECH Horizontal Modular Storage (NUHOMS) system. The Oconee version of the NUHOMS system takes advantage of burnup credit in demonstrating criticality safety. The burnup credit criticality analysis was performed by Duke Power Company's Design Engineering Department. This paper was prepared to summarize the criticality control design features employed in the Oconee NUHOMS-24P DSC basket and to describe the incentives for pursuing a burnup credit design. Principal criticality design parameters, criteria, and analysis methodology are also presented

  7. Status of burnup credit implementation in Switzerland

    International Nuclear Information System (INIS)

    Grimm, P.

    1998-01-01

    Burnup credit is currently not used for the storage of spent fuel in the reactor pools in Switzerland, but credit is taken for integral burnable absorbers. Interest exists to take credit of burnup in future for the storage in a central away-from-reactor facility presently under construction. For spent fuel transports to foreign reprocessing plants the regulations of the receiving countries must be applied in addition to the Swiss licensing criteria. Burnup credit has been applied by one Swiss PWR utility for such transports in a consistent manner with the licensing practice in the receiving countries. Measurements of reactivity worths of small spent fuel samples in a Swiss zero-power research reactor are at an early stage of planning. (author)

  8. Implementation of burnup in FERM nodal computer code

    International Nuclear Information System (INIS)

    Yoriyaz, H.; Nakata, H.

    1986-01-01

    In this work a spatial burnup scheme and feedback effects has been implemented into the FERM [1] ('Finite Element Response Matrix') program. The spatially dependent neutronic parameters have been considered in three levels: zonewise calculation, assemblywise calculation and pointwise calculation. The results have been compared with the results obtained by CITATION [2] program and showed that the processing time in the FERM code has been hundred of times shorter and no significant difference has been observed in the assembly average power distribution. (Author) [pt

  9. Nuclear fuel burn-up economy

    International Nuclear Information System (INIS)

    Matausek, M.

    1984-01-01

    In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

  10. The octopus burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

    1996-09-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  11. The OCTOPUS burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

  12. The octopus burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

    1996-01-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  13. The OCTOPUS burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

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

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1994-01-01

    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

  15. A new approach to make collapsed cross section for burnup calculation of subcritical system

    International Nuclear Information System (INIS)

    Matsunaka, Masayuki; Kondo, Keitaro; Miyamaru, Hiroyuki; Murata, Isao

    2008-01-01

    A general-purpose transport and burnup code system for precise analysis of subcritical reactors like a fusion-fission (FF) hybrid reactor was developed and used for analyzing their performance. The FF hybrid reactor is a subcritical system, which has a concept of fusion reactor with a blanket region containing nuclear fuel and has been under discussion by author's group for years because the present burnup calculation system mainly consists of a general-purpose Monte Carlo code MCNP-4B, a point burnup code ORIGEN2. JENDL-3.3 pointwise cross section library and JENDL Activation Cross Section File 96 were used as base cross section libraries to make group constant for burnup calculation. A new method has been proposed to make group constant for the burnup calculation as accurate as possible directly using output data of the neutron transport calculation by MCNP and evaluated nuclear data libraries. This method is strict and a general procedure to make one group cross sections in Monte Carlo calculations, while it takes very long computation time. Some speed-up techniques were discussed for the present group constant making process so as to decrease calculation time. Adoption of postprocessing to make group constant improved the calculation accuracy because of increasing number of cross sections to be updated in each burnup cycle. The present calculation system is capable of performing neutronics analysis of subcritical reactors more precise than our previous one. However, at the moment, it still takes long computation time to make group constants. Further speed-up techniques are now under investigation so as to apply the present system to neutronics design analysis for various subcritical systems. (author)

  16. Effect of core burnup on the dynamic behavior of fast reactors

    International Nuclear Information System (INIS)

    Ilberg, D.; Saphier, D.; Yiftah, S.

    1977-01-01

    Performance of a dynamic analysis, taking burnup changes into account, requires fission-product nuclear data of relatively small uncertainty, suitable burnup calculation models, and dynamic computer programs. These were prepared and used with the following results: (1) Significant changes in static and dynamic parameters were observed when investigating the effect of burnup. These changes were found to be larger than differences introduced by the uncertainty of the fission-product nuclear data. (2) A one-dimensional burnup computer program was prepared. It was found that a burnup model based on the generalized radioactive decay scheme is suitable for accurate fast reactor calculations. (3) Space-time dynamic calculations of fast reactors having different burnup levels were performed. The stability difference between ''clean'' and high burnup cores is greater when local rather than uniform perturbations are inserted along the entire core length. The magnitude by which the ''end-of-life'' core increases the transient excursion over that of the clean core depends on the particular region in which the perturbation is inserted. The end-of-life core will magnify the transient excursion more than the clean core whenever the perturbation is inserted into a region having a higher adjoint flux level than that of the clean core. However, when a reactor safety system operates successfully, the difference in the temperature transient of the clean and end-of-life cores will be relatively small. It is suggested that only the analysis of large local perturbations be performed for end-of-life cores as well as for clean cores in the safety evaluation of fast reactors

  17. Non destructive assay of nuclear LEU spent fuels for burnup credit application

    International Nuclear Information System (INIS)

    Lebrun, A.; Bignan, G.

    2001-01-01

    Criticality safety analysis devoted to spent fuel storage and transportation has to be conservative in order to be sure no accident will ever happen. In the spent fuel storage field, the assumption of freshness has been used to achieve the conservative aspect of criticality safety procedures. Nevertheless, after being irradiated in a reactor core, the fuel elements have obviously lost part of their original reactivity. The concept of taking into account this reactivity loss in criticality safety analysis is known as Burnup credit. To be used, Burnup credit involves obtaining evidence of the reactivity loss with a Burnup measurement. Many non destructive assays (NDA) based on neutron as well as on gamma ray emissions are devoted to spent fuel characterization. Heavy nuclei that compose the fuels are modified during irradiation and cooling. Some of them emit neutrons spontaneously and the link to Burnup is a power link. As a result, burn-up determination with passive neutron measurement is extremely accurate. Some gamma emitters also have interesting properties in order to characterize spent fuels but the convenience of the gamma spectrometric methods is very dependent on characteristics of spent fuel. In addition, contrary to the neutron emission, the gamma signal is mostly representative of the peripheral rods of the fuels. Two devices based on neutron methods but combining different NDA methods which have been studied in the past are described in detail: 1. The PYTHON device is a combination of a passive neutron measurement, a collimated total gamma measurement, and an online depletion code. This device, which has been used in several Nuclear Power Plants in western Europe, gives the average Burnup within a 5% uncertainty and also the extremity Burnup, 2. The NAJA device is an automatic device that involves three nuclear methods and an online depletion code. It is designed to cover the whole fuel assembly panel (Active Neutron Interrogation, Passive Neutron

  18. Fuel analysis code FAIR and its high burnup modelling capabilities

    International Nuclear Information System (INIS)

    Prasad, P.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1995-01-01

    A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs

  19. Burnup calculation for a tokamak commercial hybrid reactor

    International Nuclear Information System (INIS)

    Feng Kaiming; Xie Zhongyou

    1990-08-01

    A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

  20. Optimum burnup of BAEC TRIGA research reactor

    International Nuclear Information System (INIS)

    Lyric, Zoairia Idris; Mahmood, Mohammad Sayem; Motalab, Mohammad Abdul; Khan, Jahirul Haque

    2013-01-01

    Highlights: ► Optimum loading scheme for BAEC TRIGA core is out-to-in loading with 10 fuels/cycle starting with 5 for the first reload. ► The discharge burnup ranges from 17% to 24% of U235 per fuel element for full power (3 MW) operation. ► Optimum extension of operating core life is 100 MWD per reload cycle. - Abstract: The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has been operating since 1986 without any reshuffling or reloading yet. Optimum fuel burnup strategy has been investigated for the present BAEC TRIGA core, where three out-to-in loading schemes have been inspected in terms of core life extension, burnup economy and safety. In considering different schemes of fuel loading, optimization has been searched by only varying the number of fuels discharged and loaded. A cost function has been defined and evaluated based on the calculated core life and fuel load and discharge. The optimum loading scheme has been identified for the TRIGA core, the outside-to-inside fuel loading with ten fuels for each cycle starting with five fuels for the first reload. The discharge burnup has been found ranging from 17% to 24% of U235 per fuel element and optimum extension of core operating life is 100 MWD for each loading cycle. This study will contribute to the in-core fuel management of TRIGA reactor

  1. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Burnup credit effect on proposed cask payloads

    International Nuclear Information System (INIS)

    Hall, I.K.

    1989-01-01

    The purpose of the Cask Systems Development Program (CSDP) is to develop a variety of cask systems which will allow safe and economical movement of commercial spent nuclear fuel and high-level waste from the generator to the Federal repository or Monitored Retrievable Storage (MRS) facility. Program schedule objectives for the initial phase of the CSDP include the development of certified spent fuel cask systems by 1995 to support Office of Civilian Radioactive Waste Management shipments from the utilities beginning in the late 1990s. Forty-nine proposals for developing a family of spent fuel casks were received and comparisons made. General conclusions that can be drawn from the comparisons are that (1) the new generation of casks will have substantially increased payloads in comparison to current casks, and (2) an even greater payload increase may be achievable with burnup credit. The ranges in the payload estimates do not allow a precise separation of the payload increase attributable to the proposed allowance of fuel burnup credit, as compared wilt the no-burnup-credit case. The beneficial effects of cask payload increases on overall costs and risks of transporting spent fuel are significant; therefore further work aimed toward taking advantage of burnup credit is warranted

  3. WWER-1000 Burnup Credit Benchmark (CB5)

    International Nuclear Information System (INIS)

    Manolova, M.A.

    2002-01-01

    In the paper the specification of WWER-1000 Burnup Credit Benchmark first phase (depletion calculations), given. The second phase - criticality calculations for the WWER-1000 fuel pin cell, will be given after the evaluation of the results, obtained at the first phase. The proposed benchmark is a continuation of the WWER benchmark activities in this field (Author)

  4. Whole core burnup calculations using 'MCNP'

    International Nuclear Information System (INIS)

    Haran, O.; Shaham, Y.

    1996-01-01

    Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors)

  5. Whole core burnup calculations using `MCNP`

    Energy Technology Data Exchange (ETDEWEB)

    Haran, O; Shaham, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    Core parameters such as the reactivity, the power distribution and different reactivity coefficients calculated in simulations play an important role in the nuclear reactor handling. Operational safety margins are decided upon, based on the calculated parameters. Thus, the ability to accurately calculate those parameters is of uppermost importance. Such ability exists for fresh cores, using the Monte-Carlo method. The change in the core parameters that results from the core burnup is nowadays calculated within transport codes that simplifies the transport process by using approximations such as the diffusion approximation. The inaccuracy in the burned core parameters arising from the use of such approximations is hard to quantify, leading to an increased gap between the operational routines and the safety limits. A Monte Carlo transport code that caries out accurate static calculations in three dimensional geometries using continuous-energy neutron cross-section data such as the MCNP can be used to generate accurate reaction rates for burnup purposes. Monte Carlo method is statistical by nature, so that the reaction rates calculated will be accurate only to a certain known extent. The purpose of this work was to create a burnup routine that uses the capabilities of the Monte Carlo based MCNP code. It should be noted that burnup using Monte Carlo has been reported in the literatures, but this work is the result of an independent effort (authors).

  6. Burnup calculations using Monte Carlo method

    International Nuclear Information System (INIS)

    Ghosh, Biplab; Degweker, S.B.

    2009-01-01

    In the recent years, interest in burnup calculations using Monte Carlo methods has gained momentum. Previous burn up codes have used multigroup transport theory based calculations followed by diffusion theory based core calculations for the neutronic portion of codes. The transport theory methods invariably make approximations with regard to treatment of the energy and angle variables involved in scattering, besides approximations related to geometry simplification. Cell homogenisation to produce diffusion, theory parameters adds to these approximations. Moreover, while diffusion theory works for most reactors, it does not produce accurate results in systems that have strong gradients, strong absorbers or large voids. Also, diffusion theory codes are geometry limited (rectangular, hexagonal, cylindrical, and spherical coordinates). Monte Carlo methods are ideal to solve very heterogeneous reactors and/or lattices/assemblies in which considerable burnable poisons are used. The key feature of this approach is that Monte Carlo methods permit essentially 'exact' modeling of all geometrical detail, without resort to ene and spatial homogenization of neutron cross sections. Monte Carlo method would also be better for in Accelerator Driven Systems (ADS) which could have strong gradients due to the external source and a sub-critical assembly. To meet the demand for an accurate burnup code, we have developed a Monte Carlo burnup calculation code system in which Monte Carlo neutron transport code is coupled with a versatile code (McBurn) for calculating the buildup and decay of nuclides in nuclear materials. McBurn is developed from scratch by the authors. In this article we will discuss our effort in developing the continuous energy Monte Carlo burn-up code, McBurn. McBurn is intended for entire reactor core as well as for unit cells and assemblies. Generally, McBurn can do burnup of any geometrical system which can be handled by the underlying Monte Carlo transport code

  7. Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

    International Nuclear Information System (INIS)

    Bowman, S.M.

    1991-04-01

    To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as ''burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs

  8. Effect of core configuration on the burnup calculations of MTR research reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Amin, E.H.; Sakr, A.M.

    2014-01-01

    Highlights: • 3D burn-up calculations of MTR-type research reactor were performed. Examination of the effect of control rod pattern on power density and neutron flux distributions is presented. • The calculations are performed using the MTR P C package and the programs (WIMS and CITVAP). • An empirical formula was generated for every fuel element type, to correlate irradiation to burn-up. - Abstract: In the present paper, three-dimensional burn-up calculations were performed using different patterns of control rods, in order to examine their effect on power density and neutron flux distributions through out the entire core and hence on the local burn-up distribution. These different cores burn-up calculations are carried out for an operating cycle equivalent to 15 Full Power Days (FPDs), with a power rating of 22 MW. Calculations were performed using an example of a typical research reactor of MTR-type using the internationally known computer codes’ package “MTR P C system”, using the cell calculation transport code WIMS-D4 with 12 energy groups and the core calculation diffusion code CITVAP with 5 energy groups. A depletion study was done and the effects on the research reactor fuel (U-235) were performed. The burn-up percentage (B.U.%) curves for every fuel element type were drawn versus irradiation (MWD/TE). Then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Charts of power density and neutron flux distribution for each core were plotted at different sections of each fuel element of the reactor core. Then a complete discussion and analysis of these curves are performed with comparison between the different core configurations, illustrating the effect of insertion or extraction of either of the four control rods directly on the neutron flux and consequently on the power distribution and burn-up. A detailed study of fuel burn-up gives detailed insight on the different B.U.% calculations

  9. Impacts of SNF burnup credit on the shipment capability of the GA-4 cask

    International Nuclear Information System (INIS)

    Mobasheran, A.S.; Lake, W.; Richardson, J.

    1996-01-01

    Scoping analyses were performed to determine the impacts of two different levels of burnup credit and two different spent fuel pickup rates on the shipment capability and the minimum fleet size of the GA-4 cask. The analyses involved developing loading curves for the GA-4 cask based on the actinide-only and principal-isotope burnup credit considerations. The analyses also involved examination of the spent nuclear fuel assembly population at nine reactor sites and categorization of the assemblies in accordance with the loading restrictions imposed. The results revealed that for the nine sites considered, depending on the level of burnup credit and the pickup rate assumed, the total savings in shipment and cask fleet costs (1994 dollars) can range from $55 million to $74 million

  10. Isocrit: a burnup credit tool for spent fuel pool storage calculations - 333

    International Nuclear Information System (INIS)

    Kucukboyaci, V.N.; Marshall, W.J.

    2010-01-01

    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 up-rate, exit temperature changes, etc) with a quick turnaround. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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

  12. Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

    International Nuclear Information System (INIS)

    Ozdemir, Levent; Acar, Banu Bulut; Zabunoglu, Okan H.

    2011-01-01

    When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239 Pu and 241 Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.

  13. Burnup credit activities being conducted in the United States

    International Nuclear Information System (INIS)

    Lake, W.

    1998-01-01

    The paper describes burnup credit activities being conducted in the U.S. where burnup credit is either being used or being planned to be used for storage, transport, and disposal of spent nuclear fuel. Currently approved uses of burnup credit are for wet storage of PWR fuel. For dry storage of spent PWR fuel, burnup credit is used to supplement a principle of moderator exclusion. These storage applications have been pursued by the private sector. The Department of Energy (DOE) which is an organization of the U.S. Federal government is seeking approval for burnup credit for transport and disposal applications. For transport of spent fuel, regulatory review of an actinide-only PWR burnup credit method is now being conducted. A request by DOE for regulatory review of actinide and fission product burnup credit for disposal of spent BWR and PWR fuel is scheduled to occur in 1998. (author)

  14. Core burn-up calculation method of JRR-3

    International Nuclear Information System (INIS)

    Kato, Tomoaki; Yamashita, Kiyonobu

    2007-01-01

    SRAC code system is utilized for core burn-up calculation of JRR-3. SRAC code system includes calculation modules such as PIJ, PIJBURN, ANISN and CITATION for making effective cross section and calculation modules such as COREBN and HIST for core burn-up calculation. As for calculation method for JRR-3, PIJBURN (Cell burn-up calculation module) is used for making effective cross section of fuel region at each burn-up step. PIJ, ANISN and CITATION are used for making effective cross section of non-fuel region. COREBN and HIST is used for core burn-up calculation and fuel management. This paper presents details of NRR-3 core burn-up calculation. FNCA Participating countries are expected to carry out core burn-up calculation of domestic research reactor by SRAC code system by utilizing the information of this paper. (author)

  15. Automated generation of burnup chain for reactor analysis applications

    International Nuclear Information System (INIS)

    Tran, Viet-Phu; Tran, Hoai-Nam; Yamamoto, Akio; Endo, Tomohiro

    2017-01-01

    This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO_2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

  16. Automated generation of burnup chain for reactor analysis applications

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Viet-Phu [VINATOM, Hanoi (Viet Nam). Inst. for Nuclear Science and Technology; Tran, Hoai-Nam [Duy Tan Univ., Da Nang (Viet Nam). Inst. of Research and Development; Yamamoto, Akio; Endo, Tomohiro [Nagoya Univ., Nagoya-shi (Japan). Dept. of Materials, Physics and Energy Engineering

    2017-05-15

    This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO{sub 2} and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

  17. Current studies related to the use of burnup credit in France

    International Nuclear Information System (INIS)

    Raby, Jerome; Lavarenne, Caroline; Barreau, Anne; Riffard, Cecile; Roque, Benedicte; Bioux, Philippe; Doucet, Michel; Guillou, Eric; Leka, Georges; Toubon, Herve

    2003-01-01

    In order to avoid criticality risks, a large number of facilities using spent fuels have been designed considering the fuel as fresh. This choice has obviously led to considerable safety margins. In the early 80's, a method was accepted by the French Safety Authorities allowing to consider the changes in the fuel composition during the depletion with some very pessimistic hypothesis: only actinides were considered and the amount of burnup used in the studies was equal to the mean burnup in the 50-least-irradiated centimeters. As many facilities still want to optimize their processes (e.g. transportation, storage, fuel reprocessing), the main companies involved in the French nuclear industry, researchers and IRSN set up a Working Group in order to study the way burnup could be taken into account in the criticality calculations, considering some fission products and a more realistic axial profile of burnup. The first of this article introduces the current French method used to take burnup into account in the criticality studies. The second part is devoted to the studies achieved by the Working Group to improve this method, especially concerning the consideration of the neutron absorption of some fission products and of an axial profile of burnup: for that purpose, some results are presented related to the steps of the process like the depletion calculations, the definition of an axial profile and the criticality calculation. In the third part, some results (keff) obtained with fission products and an axial profile are compared to those obtained with the current one. The conclusions presented are related to the present state of knowledge and may differ from the final conclusions of the Working Group. (author)

  18. Development of a one-group cross section data base of the ORIGEN2 computer code for research reactor applications

    International Nuclear Information System (INIS)

    Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai; Hwang, Won Guk

    1992-01-01

    A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author)

  19. Development of a one-group cross section data base of the ORIGEN2 computer code for research reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Hwang, Won Guk [Kyung Hee University, Seoul (Korea, Republic of)

    1992-03-01

    A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author).

  20. Reconstruction of pin burnup characteristics from nodal calculations in hexagonal geometry

    International Nuclear Information System (INIS)

    Yang, W.S.; Finck, P.J.; Khalil, H.S.

    1990-01-01

    A reconstruction method has been developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 code system. Intra-modal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method has been tested by performing several numerical benchmark calculations and by comparing predicted local burnups to values measured for experimental assemblies in EBR-11. The results indicate that the reconstruction methods are quite accurate, yielding maximum errors in power and nuclide densities that are less than 2% for driver assemblies and typically less than 5% for blanket assemblies. 14 refs., 2 figs., 5 tabs

  1. Higher order methods for burnup calculations with Bateman solutions

    International Nuclear Information System (INIS)

    Isotalo, A.E.; Aarnio, P.A.

    2011-01-01

    Highlights: → Average microscopic reaction rates need to be estimated at each step. → Traditional predictor-corrector methods use zeroth and first order predictions. → Increasing predictor order greatly improves results. → Increasing corrector order does not improve results. - Abstract: A group of methods for burnup calculations solves the changes in material compositions by evaluating an explicit solution to the Bateman equations with constant microscopic reaction rates. This requires predicting representative averages for the one-group cross-sections and flux during each step, which is usually done using zeroth and first order predictions for their time development in a predictor-corrector calculation. In this paper we present the results of using linear, rather than constant, extrapolation on the predictor and quadratic, rather than linear, interpolation on the corrector. Both of these are done by using data from the previous step, and thus do not affect the stepwise running time. The methods were tested by implementing them into the reactor physics code Serpent and comparing the results from four test cases to accurate reference results obtained with very short steps. Linear extrapolation greatly improved results for thermal spectra and should be preferred over the constant one currently used in all Bateman solution based burnup calculations. The effects of using quadratic interpolation on the corrector were, on the other hand, predominantly negative, although not enough so to conclusively decide between the linear and quadratic variants.

  2. Fission product margin in burnup credit analyses

    International Nuclear Information System (INIS)

    Finck, P.J.; Stenberg, C.G.

    1998-01-01

    The US Department of Energy (DOE) is currently working toward the licensing of a methodology for using actinide-only burnup credit for the transportation of spent nuclear fuel (SNF). Important margins are built into this methodology. By using comparisons with a representative experimental database to determine bias factors, the methodology ensures that actinide concentrations and worths are estimated conservatively; furthermore, the negative net reactivity of certain actinides and all fission products (FPs) is not taken into account, thus providing additional margin. A future step of DOE's effort might aim at establishing an actinide and FP burnup credit methodology. The objective of this work is to establish the uncertainty to be applied to the total FP worth in SNF. This will serve two ends. First, it will support the current actinide-only methodology by demonstrating the margin available from FPs. Second, it will identify the major contributions to the uncertainty and help set priorities for future work

  3. Study of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    Pavelescu, M.; Borza, M.

    1975-01-01

    The authors approach theoretical treatment of isotopic composition changement for nuclear fuel in nuclear reactors. They show the difficulty of exhaustive treatment of burn-up problems and introduce the principal simplifying principles. Due to these principles they write and solve analytically the evolution equations of the concentration for the principal nuclides both in the case of fast and thermal reactors. Finally, they expose and comment the results obtained in the case of a power fast reactor. (author)

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

    International Nuclear Information System (INIS)

    Wiesenack, W.

    1996-01-01

    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

  5. Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

    International Nuclear Information System (INIS)

    Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L.; Saito, M.

    2003-01-01

    Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, 237 Np, 238 Pu, 231 Pa, 232 U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations

  6. Fuel cycles with high fuel burn-up: analysis of reactivity coefficients

    Energy Technology Data Exchange (ETDEWEB)

    Kryuchkov, E.F.; Shmelev, A.N.; Ternovykh, M.J.; Tikhomirov, G.V.; Jinhong, L. [Moscow Engineering Physics Institute (State University) (Russian Federation); Saito, M. [Tokyo Institute of Technology (Japan)

    2003-07-01

    Fuel cycles of light-water reactors (LWR) with high fuel burn-up (above 100 MWd/kg), as a rule, involve large amounts of fissionable materials. It leads to forming the neutron spectrum harder than that in traditional LWR. Change of neutron spectrum and significant amount of non-traditional isotopes (for example, {sup 237}Np, {sup 238}Pu, {sup 231}Pa, {sup 232}U) in such fuel compositions can alter substantially reactivity coefficients as compared with traditional uranium-based fuel. The present work addresses the fuel cycles with high fuel burn-up which are based on Th-Pa-U and U-Np-Pu fuel compositions. Numerical analyses are carried out to determine effective neutron multiplication factor and void reactivity coefficient (VRC) for different values of fuel burn-up and different lattice parameters. The algorithm is proposed for analysis of isotopes contribution to these coefficients. Various ways are considered to upgrade safety of nuclear fuel cycles with high fuel burn-up. So, the results obtained in this study have demonstrated that: -1) Non-traditional fuel compositions developed for achievement of high fuel burn-up in LWR can possess positive values of reactivity coefficients that is unacceptable from the reactor operation safety point of view; -2) The lattice pitch of traditional LWR is not optimal for non-traditional fuel compositions, the increased value of the lattice pitch leads to larger value of initial reactivity margin and provides negative VRC within sufficiently broad range of coolant density; -3) Fuel burn-up has an insignificant effect on VRC dependence on coolant density, so, the measures undertaken to suppress positive VRC of fresh fuel will be effective for partially burnt-up fuel compositions also and; -4) Increase of LWR core height and introduction of additional moderators into the fuel lattice can be used as the ways to reach negative VRC values for full range of possible coolant density variations.

  7. Conceptual cask design with burnup credit

    International Nuclear Information System (INIS)

    Lee, Seong Hee; Ahn, Joon Gi; Hwang, Hae Ryong

    2003-01-01

    Conceptual design has been performed for a spent fuel transport cask with burnup credit and a neutron-absorbing material to maximize transportation capacity. Both fresh and burned fuel are assumed to be stored in the cask and boral and borated stainless steel are selected for the neutron-absorbing materials. Three different sizes of cask with typical 14, 21 and 52 PWR fuel assemblies are modeled and analyzed with the SCALE 4.4 code system. In this analysis, the biases and uncertainties through validation calculations for both isotopic predictions and criticality calculation for the spent fuel have been taken into account. All of the reactor operating parameters, such as moderator density, soluble boron concentration, fuel temperature, specific power, and operating history, have been selected in a conservative way for the criticality analysis. Two different burnup credit loading curves are developed for boral and borated stainless steel absorbing materials. It is concluded that the spent fuel transport cask design with burnup credit is feasible and is expected to increase cask payloads. (author)

  8. A Preliminary Investigation of a Randomized Dependent Group Contingency for Hallway Transitions

    Science.gov (United States)

    Deshais, Meghan A.; Fisher, Alyssa B.; Kahng, SungWoo

    2018-01-01

    We conducted a preliminary investigation of a randomized dependent group contingency to decrease disruptive behavior during hallway transitions. Two first-graders, identified by their classroom teacher, participated in this study. A multiple baseline across transitions was used to evaluate the effects of the randomized dependent group contingency…

  9. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1996-05-01

    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. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports

  10. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1996-05-01

    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. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  11. Biases and statistical errors in Monte Carlo burnup calculations: an unbiased stochastic scheme to solve Boltzmann/Bateman coupled equations

    International Nuclear Information System (INIS)

    Dumonteil, E.; Diop, C.M.

    2011-01-01

    External linking scripts between Monte Carlo transport codes and burnup codes, and complete integration of burnup capability into Monte Carlo transport codes, have been or are currently being developed. Monte Carlo linked burnup methodologies may serve as an excellent benchmark for new deterministic burnup codes used for advanced systems; however, there are some instances where deterministic methodologies break down (i.e., heavily angularly biased systems containing exotic materials without proper group structure) and Monte Carlo burn up may serve as an actual design tool. Therefore, researchers are also developing these capabilities in order to examine complex, three-dimensional exotic material systems that do not contain benchmark data. Providing a reference scheme implies being able to associate statistical errors to any neutronic value of interest like k(eff), reaction rates, fluxes, etc. Usually in Monte Carlo, standard deviations are associated with a particular value by performing different independent and identical simulations (also referred to as 'cycles', 'batches', or 'replicas'), but this is only valid if the calculation itself is not biased. And, as will be shown in this paper, there is a bias in the methodology that consists of coupling transport and depletion codes because Bateman equations are not linear functions of the fluxes or of the reaction rates (those quantities being always measured with an uncertainty). Therefore, we have to quantify and correct this bias. This will be achieved by deriving an unbiased minimum variance estimator of a matrix exponential function of a normal mean. The result is then used to propose a reference scheme to solve Boltzmann/Bateman coupled equations, thanks to Monte Carlo transport codes. Numerical tests will be performed with an ad hoc Monte Carlo code on a very simple depletion case and will be compared to the theoretical results obtained with the reference scheme. Finally, the statistical error propagation

  12. Evaluation of burnup characteristics and energy deposition during NSRR pulse irradiation tests on irradiated BWR fuels

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio

    2000-11-01

    Pulse irradiation tests of irradiated fuel are performed in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under Reactivity Initiated Accident Conditions (RIA). The severity of the RIA is represented by energy deposition or peak fuel enthalpy during the power excursion. In case of the irradiated fuel tests, the energy deposition varies depending both on the amounts and distribution of residual fissile and neutron absorbing fission products generated during the base irradiation. Thus, proper fuel burnup characterization, especially for low enriched commercial fuels, is important, because plutonium (Pu) takes a large part of fissile and its generation depends on the neutron spectrum during the base irradiation. Fuel burnup calculations were conducted with ORIGEN2, RODBURN and SWAT codes for the BWR fuels tested in the NSRR. The calculation results were compared with the measured isotope concentrations and used for the NSRR neutron calculations to evaluate energy depositions of the test fuel. The comparison of the code calculations and the measurements revealed that the neutron spectrum change due to difference in void fraction altered Pu generation and energy deposition in the NSRR tests considerably. With the properly evaluated neutron spectrum, the combined burnup and NSRR neutron calculation gave reasonably good evaluation of the energy deposition. The calculations provided radial distributions of the fission product accumulation during the base irradiation and power distribution during the NSRR pulse irradiation, which were important for the evaluation of both burnup characteristics and fission gas release behavior. (author)

  13. Time step length versus efficiency of Monte Carlo burnup calculations

    International Nuclear Information System (INIS)

    Dufek, Jan; Valtavirta, Ville

    2014-01-01

    Highlights: • Time step length largely affects efficiency of MC burnup calculations. • Efficiency of MC burnup calculations improves with decreasing time step length. • Results were obtained from SIE-based Monte Carlo burnup calculations. - Abstract: We demonstrate that efficiency of Monte Carlo burnup calculations can be largely affected by the selected time step length. This study employs the stochastic implicit Euler based coupling scheme for Monte Carlo burnup calculations that performs a number of inner iteration steps within each time step. In a series of calculations, we vary the time step length and the number of inner iteration steps; the results suggest that Monte Carlo burnup calculations get more efficient as the time step length is reduced. More time steps must be simulated as they get shorter; however, this is more than compensated by the decrease in computing cost per time step needed for achieving a certain accuracy

  14. Burnup credit applications in a high-capacity truck cask

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1993-01-01

    The use of burnup credit in the criticality safety analysis of the GA-4 Cask increases the cask's capacity from three spent fuel assemblies to four, resulting in reduced public and occupational risk and reduced life cycle costs. GA's criticality calculations for burnup credit, including the associated uncertainties and analytical bias, establish the minimum burnup required as a function of initial enrichment to maintain K eff ≤ 0.95 under any conceivable condition. The minimum burnup requirement as a function of initial enrichment has been determined to be 15,000 MWd/MTU for 3.5 wt% U-235 fuel, 20,000 MWd/MTU for 4.0 wt% U-235 fuel and 25,000 MWd/MTU for 4.5 wt% U-235 fuel. The minimum burnup requirement as a function of enrichment is well below the typical burnup levels seen in the current and projected spent fuel inventory. (J.P.N.)

  15. Implementation of burnup credit in spent fuel management systems

    International Nuclear Information System (INIS)

    Dyck, H.P.

    2001-01-01

    Improved calculational methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservatism while maintaining an adequate safety margin. The motivation for using burnup credit in criticality safety applications is based on economic considerations and additional benefits contributing to public health and safety and resource conservation. Interest in the implementation of burnup credit has been shown by many countries. In 1997, the International Atomic Energy Agency (IAEA) started a task to monitor the implementation of burnup credit in spent fuel management systems, to provide a forum to exchange information, to discuss the matter and to gather and disseminate information on the status of national practices of burnup credit implementation in the Member States. The task addresses current and future aspects of burnup credit. This task was continued during the following years. (author)

  16. Theory analysis and simple calculation of travelling wave burnup scheme

    International Nuclear Information System (INIS)

    Zhang Jian; Yu Hong; Gang Zhi

    2012-01-01

    Travelling wave burnup scheme is a new burnup scheme that breeds fuel locally just before it burns. Based on the preliminary theory analysis, the physical imagine was found. Through the calculation of a R-z cylinder travelling wave reactor core with ERANOS code system, the basic physical characteristics of this new burnup scheme were concluded. The results show that travelling wave reactor is feasible in physics, and there are some good features in the reactor physics. (authors)

  17. Burnup calculation in microcells of high conversion reactors

    International Nuclear Information System (INIS)

    Gomez, S.E.; Salvatore, M.; Patino, N.E.; Abbate, M.J.

    1991-01-01

    The development of high converter reactors (HCR) requires careful burnup calculations because their main goals are reach high discharge burnup levels (Up to 50 GWd/T) and a close to one conversion ratio. Then, it is necessary a revision of design elements used for this type of calculation. In this work, a burnup module (BUM) developed in order to use nuclear data directly from evaluated data files is presented; these was included in the AMPX system. (author)

  18. Fuel performance at high burnup for water reactors

    International Nuclear Information System (INIS)

    1991-02-01

    The present meeting was scheduled by the International Atomic Energy Agency, upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The purpose of this meeting was to review the ''state-of-the-art'' in the area of Fuel Performance at High Burnup for Water Reactors. Previous IAEA meetings on this topic were held in Mol in 1981 and 1984 and on related topics in Stockholm and Lyon in 1987. Fifty-five participants from 16 countries and two international organizations attended the meeting and 28 papers were presented and discussed. The papers were presented in five sub-sessions and during the meeting, working groups composed of the session chairmen and paper authors prepared the summary of each session with conclusions and recommendations for future work. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  19. Technical and economic limits to fuel burnup extension. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2002-07-01

    For many years, the increase of efficiency in the production of nuclear electricity has been an economic challenge in many countries which have developed this kind of energy. The increase of fuel burnup leads to a reduction in the volume of spent fuel discharged to longer fuel cycles in the reactor, which means bigger availability and capacity factors. After having increased the authorized burnup in plants, developing new alloys capable of resisting high burnup, and having accumulated data on fuel evolution with burnup, it has become necessary to establish the limitations which could be imposed by the physical evolution of the fuel, influencing fuel management, neutron properties, reprocessing or, more generally, the management of waste and irradiated fuels. It is also necessary to verify whether the benefits of lower electricity costs would not be offset by an increase in fuel management costs. The main questions are: Are technical and economic limits to the increasing of fuel burnup in parallel? Can we envisage nowadays the hardest limitation in some of these areas? Which are the main points to be solved from the technical point of view? Is this effort worthwhile considering the economy of the cycle? To which extent? For these reasons, the IAEA, following a recommendation by the International Working Group on Fuel Performance and Technology, held a Technical Committee Meeting on Technical and Economic Limits to Fuel Burnup Extension. The purpose of this meeting was to provide an international forum to review the evolution of fuel properties at increased burnup in order to estimate the limitations both from a physical and an economic point of view. The meeting was therefore divided into two parts. The first part, focusing on technical limits, was devoted to the improvement of the fuel element, such as fission gas release (FGR), RIM effect, cladding, etc. and the fabrication, core management, spent fuel and reprocessing. Eighteen related papers were presented which

  20. Time resolved measurements of triton burnup in JET plasmas

    International Nuclear Information System (INIS)

    Conroy, S.; Jarvis, O.N.; Sadler, G.; Huxtable, G.B.

    1988-01-01

    Triton production from one branch of the deuteron-deuteron fusion reaction is routinely measured at 6 ms time intervals in JET plasma discharges by recording the 2.5 MeV neutrons produced in the other branch using a set of calibrated fission chambers. The burnup of the tritons is measured by detecting the 14 MeV t-d neutrons with a 0.2 cm 3 Si(Li) diode. The 2.5 MeV neutron flux can be used in a simple time dependent calculation based on classical slowing-down theory to predict the 14 MeV neutron flux. The measured flux and the triton slowing-down time are systematically lower than the values estimated from the key plasma parameters but the differences are within the experimental errors. (author). 19 refs, 8 figs

  1. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Henriquez, C; Navarro, G; Pereda, C

    2000-01-01

    Some time ago the CCHEN authorities decided to produce a set of 50 low enrichment fuel elements. These elements were produced in the PEC (Fuel Elements Plant), located at CCHEN offices in Lo Aguirre. These new fuel elements have basically the same geometrical characteristics of previous ones, which were British and made with raw material from the U.S. The principal differences between our fuel elements and the British ones is the density of fissile material, U-235, which was increased to compensate the reduction in enrichment. Last year, the Fuel Elements Plant (PEC) delivered the shipment's first four (4) fuel elements, called leaders, to the RECH1. A test element was delivered too, and the complete set was introduced into the reactor's nucleus, following the normal routine, but performing a special follow-up on their behavior inside the nucleus. This experimental element has only one outside fuel plate, and the remaining (15) structural plates are aluminum. In order to study the burnup, the test element was taken out of the nucleus, in mid- November 1999, and left to decay until June 2000, when it was moved to the laboratory (High Activity Cell), to start the burnup measurements, with a gamma spectroscopy system. This work aims to show the results of these measurements and in addition to meet the following objectives: (a) Visual test of the plate's general condition; (b) Sipping test of fission products; (c) Study of burn-up distribution in the plate; (d) Check and improve the calculus algorithm; (e) Comparison of the results obtained from the spectroscopy with the ones from neutron calculus

  2. Study of the acceleration of nuclide burnup calculation using GPU with CUDA

    International Nuclear Information System (INIS)

    Okui, S.; Ohoka, Y.; Tatsumi, M.

    2009-01-01

    The computation costs of neutronics calculation code become higher as physics models and methods are complicated. The degree of them in neutronics calculation tends to be limited due to available computing power. In order to open a door to the new world, use of GPU for general purpose computing, called GPGPU, has been studied [1]. GPU has multi-threads computing mechanism enabled with multi-processors which realize mush higher performance than CPUs. NVIDIA recently released the CUDA language for general purpose computation which is a C-like programming language. It is relatively easy to learn compared to the conventional ones used for GPGPU, such as OpenGL or CG. Therefore application of GPU to the numerical calculation became much easier. In this paper, we tried to accelerate nuclide burnup calculation, which is important to predict nuclides time dependence in the core, using GPU with CUDA. We chose the 4.-order Runge-Kutta method to solve the nuclide burnup equation. The nuclide burnup calculation and the 4.-order Runge-Kutta method were suitable to the first step of introduction CUDA into numerical calculation because these consist of simple operations of matrices and vectors of single precision where actual codes were written in the C++ language. Our experimental results showed that nuclide burnup calculations with GPU have possibility of speedup by factor of 100 compared to that with CPU. (authors)

  3. Burnup characteristics of binary breeder reactors

    International Nuclear Information System (INIS)

    Dias, A.F.; Nascimento, J.A. do; Ishiguro, Y.

    1983-01-01

    Burnup calculations of a binary breeder reactor have been done for two cases of fueling. In one case the U 233 /TH fueled inner core and the Pu/U-fueled outer core have the same number of fuel assemblies. In the other case two outermost rings in the inner core are Pu/U-fueled. The second case is considered for an initial phase of thorim cycle introduction when the supply of U 233 could be limited. Results show an efficient breeding on the thorium cycle in both cases. (Author) [pt

  4. High Burnup Fuel Performance and Safety Research

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Je Keun; Lee, Chan Bok; Kim, Dae Ho (and others)

    2007-03-15

    The worldwide trend of nuclear fuel development is to develop a high burnup and high performance nuclear fuel with high economies and safety. Because the fuel performance evaluation code, INFRA, has a patent, and the superiority for prediction of fuel performance was proven through the IAEA CRP FUMEX-II program, the INFRA code can be utilized with commercial purpose in the industry. The INFRA code was provided and utilized usefully in the universities and relevant institutes domesticallly and it has been used as a reference code in the industry for the development of the intrinsic fuel rod design code.

  5. Technological and licensing challenges for high burnup fuel

    International Nuclear Information System (INIS)

    Gross, H.; Urban, P.; Fenzlein, C.

    2002-01-01

    Deregulation of electricity markets is driving electricity prices downward as well in the U.S. as in Europe. As a consequence high burnup fuel will be demanded by utilities using either the storage or the reprocessing option. At a minimum, burnups consistent with the current political enrichment limit of 5 w/o will be required for both markets.Significant progress has been achieved in the past by Siemens in meeting the demands of utilities for increased fuel burnup. The technological challenges posed by the increased burnup are mainly related to the corrosion and hydrogen pickup of the clad, the high burnup properties of the fuel and the dimensional changes of the fuel assembly structure. Clad materials with increased corrosion resistance appropriate for high burnup have been developed. The high burnup behaviour of the fuel has been extensively investigated and the decrease of thermal conductivity with burnup, the rim effect of the pellet and the increase of fission gas release with burnup can be described, with good accuracy, in fuel rod computer codes. Advanced statistical design methods have been developed and introduced. Materials with increased corrosion resistance are also helpful controlling the dimensional changes of the fuel assembly structure. In summary, most of the questions about the fuel operational behaviour and reliability in the high burnup range have been solved - some of them are still in the process of verification - or the solutions are visible. This fact is largely acknowledged by regulators too. The main licensing challenges for high burnup fuel are currently seen for accident condition analyses, especially for RIA and LOCA. (author)

  6. Visualization of fuel rod burnup analysis by Scilab

    International Nuclear Information System (INIS)

    Tsai, Chiung-Wen

    2013-01-01

    The goal of this technical note is to provide an alternative, the freeware Scilab, by which means we may construct custom GUIs and distribute them without extra constrains and cost. A post-processor has been constructed by Scilab to visualize the fuel rod burnup analysis data calculated by FRAPCON-3.4. This post-processor incorporates a graphical user interface (GUI), providing users a rapid overview of the characteristics of the numerical results with 2-D and 3-D graphs, as well as the animations of fuel temperature distribution. An assessment case input file provided by FRAPCON user group was applied to demonstrate the construction of a post-processor with GUI by object-oriented GUI tool, as well as the capability of visualization functions of Scilab

  7. Visualization of fuel rod burnup analysis by Scilab

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, Chiung-Wen, E-mail: d937121@oz.nthu.edu.tw

    2013-12-15

    The goal of this technical note is to provide an alternative, the freeware Scilab, by which means we may construct custom GUIs and distribute them without extra constrains and cost. A post-processor has been constructed by Scilab to visualize the fuel rod burnup analysis data calculated by FRAPCON-3.4. This post-processor incorporates a graphical user interface (GUI), providing users a rapid overview of the characteristics of the numerical results with 2-D and 3-D graphs, as well as the animations of fuel temperature distribution. An assessment case input file provided by FRAPCON user group was applied to demonstrate the construction of a post-processor with GUI by object-oriented GUI tool, as well as the capability of visualization functions of Scilab.

  8. Simulation of the behaviour of nuclear fuel under high burnup conditions

    International Nuclear Information System (INIS)

    Soba, Alejandro; Lemes, Martin; González, Martin Emilio; Denis, Alicia; Romero, Luis

    2014-01-01

    Highlights: • Increasing the time of nuclear fuel into reactor generates high burnup structure. • We analyze model to simulate high burnup scenarios for UO 2 nuclear fuel. • We include these models in the DIONISIO 2.0 code. • Tests of our models are in very good agreement with experimental data. • We extend the range of predictability of our code up to 60 MWd/KgU average. - Abstract: In this paper we summarize all the models included in the latest version of the DIONISIO code related to the high burnup scenario. Due to the extension of nuclear fuels permanence under irradiation, physical and chemical modifications are developed in the fuel material, especially in the external corona of the pellet. The codes devoted to simulation of the rod behaviour under irradiation need to introduce modifications and new models in order to describe those phenomena and be capable to predict the behaviour in all the range of a general pressurized water reactor. A complex group of subroutines has been included in the code in order to predict the radial distribution of power density, burnup, concentration of diverse nuclides and porosity within the pellet. The behaviour of gadolinium as burnable poison also is modelled into the code. The results of some of the simulations performed with DIONISIO are presented to show the good agreement with the data selected for the FUMEX I/II/III exercises, compiled in the NEA data bank

  9. Overview of the burnup credit activities at OECD/NEA/NSC

    International Nuclear Information System (INIS)

    Brady Raap, M.C.; Nomura, Y.; Sartori, E.

    2001-01-01

    This article summarizes activities of the OECD/NEA Burnup Credit Expert Panel, a subordinate group to the Working Party on Nuclear Criticality Safety (WPNCS). The WPNCS of the OECD/NEA coordinates and carries out work in the domain of criticality safety at the international level. Particular attention is devoted to establishing sound databases required in this area and to addressing issues of high relevance such as burnup credit. The activities of the expert panel are aimed toward improving safety and identifying economic solutions to issues concerning the back-end of the fuel cycle. The main objective of the activities of the OECD/NEA Burnup Credit Expert Panel is to demonstrate that the available criticality safety calculational tools are appropriate for application to burned fuel systems and that a reasonable safety margin can be established. The method established by the expert panel for investigating the physics and predictability of burnup credit is based on the specification and comparison of calculational benchmark problems. A wide range of fuel types, including PWR, BWR, MOX, and VVER fuels, has been or are being addressed by the expert panel. The objective and status of each of these benchmark problems is reviewed in this article. It is important to note that the focus of the expert panel is the comparison of the results submitted by each participant to assess the capability of commonly used code systems, not to quantify the physical phenomena investigated in the comparisons or to make recommendations for licensing action. (author)

  10. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  11. Device for measuring a burnup degree

    International Nuclear Information System (INIS)

    Ito, Toshiaki; Goto, Seiichiro

    1979-01-01

    Purpose: To measure the burnup degree at high efficiency and accuracy. Constitution: The outer metal wall of fuel assemblies is heated under gamma radiation with long half life gamma rays in inverse proportion to the burnup degree and issues infrared radiation in proportion to the intensity of the gamma rays. An image pick-up tube is opposed to one surface of the fuel assemblies to detect the radiated infrared rays. Since the output signal from the pick-up tube is subjected to the absorptive damping by the distance between the pick-up tube and the fuel assembly, as well as water filled in the gap therebetween, it is corrected through a main amplifier comprising a signal correction circuit composed of a characteristic section inverse to the absorption property and a characteristic section inverse to the square of the distance. The corrected output signal is displayed on a display unit such as CRT or recorded in a film or a magnetic tape. (Furukawa, Y.)

  12. Actinides burnup in a sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Pineda A, R.; Martinez C, E.; Alonso, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2017-09-15

    The burnup of actinides in a nuclear reactor is been proposed as part of an advanced nuclear fuel cycle, this process would close the fuel cycle recycling some of the radioactive material produced in the open nuclear fuel cycle. These actinides are found in the spent nuclear fuel from nuclear power reactors at the end of their burnup in the reactor. Previous studies of actinides recycling in thermal reactors show that would be possible reduce the amounts of actinides at least in 50% of the recycled amounts. in this work, the amounts of actinides that can be burned in a fast reactor is calculated, very interesting results surge from the calculations, first, the amounts of actinides generated by the fuel is higher than for thermal fuel and the composition of the actinides vector is different as in fuel for thermal reactor the main isotope is the {sup 237}Np in the fuel for fast reactor the main isotope is the {sup 241}Am, finally it is concluded that the fast reactor, also generates important amounts of waste. (Author)

  13. ABB PWR fuel design for high burnup

    International Nuclear Information System (INIS)

    Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.

    1998-01-01

    Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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 149 Sm, 151 Sm, and 155 Gd

  16. Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydın, E-mail: karahan@alum.mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States); Andrews, Nathan C., E-mail: nandrews@mit.edu [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-215, Cambridge, MA 02139 (United States)

    2013-05-15

    Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

  17. Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

    International Nuclear Information System (INIS)

    Karahan, Aydın; Andrews, Nathan C.

    2013-01-01

    Highlights: ► Improved fuel swelling models in phase structure dependent form. ► A probabilistic verification exercise for the open porosity formation threshold. ► Satisfactory validation effort for available EBR-II database. ► Ultra high burn-up behavior of U–6Zr fuel with 60% smear density fuel. -- Abstract: Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α + δ) and (β + γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented

  18. Revised SWAT. The integrated burnup calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  19. Benchmarking burnup reconstruction methods for dynamically operated research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sternat, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Charlton, William S. [Univ. of Nebraska, Lincoln, NE (United States). National Strategic Research Institute; Nichols, Theodore F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-01

    The burnup of an HEU fueled dynamically operated research reactor, the Oak Ridge Research Reactor, was experimentally reconstructed using two different analytic methodologies and a suite of signature isotopes to evaluate techniques for estimating burnup for research reactor fuel. The methods studied include using individual signature isotopes and the complete mass spectrometry spectrum to recover the sample’s burnup. The individual, or sets of, isotopes include 148Nd, 137Cs+137Ba, 139La, and 145Nd+146Nd. The storage documentation from the analyzed fuel material provided two different measures of burnup: burnup percentage and the total power generated from the assembly in MWd. When normalized to conventional units, these two references differed by 7.8% (395.42GWd/MTHM and 426.27GWd/MTHM) in the resulting burnup for the spent fuel element used in the benchmark. Among all methods being evaluated, the results were within 11.3% of either reference burnup. The results were mixed in closeness to both reference burnups; however, consistent results were achieved from all three experimental samples.

  20. Revised SWAT. The integrated burnup calculation code system

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  1. Burn-up measurements coupling gamma spectrometry and neutron measurement

    Energy Technology Data Exchange (ETDEWEB)

    Toubon, H.; Pin, P. [AREVA/CANBERRA, 1 rue des Herons, 78182 St Quentin-en-Yvelines Cedex (France); Lebrun, A. [IAEA, Wagramer Strasse 5, PO Box 100, Vienna (Austria); Oriol, L.; Saurel, N. [CEA Cadarache, 13108 Saint Paul Lez Durance Cedex (France); Gain, T. [AREVA/COGEMA Reprocessing Business Unit, La Hague, 50444 Beaumont Hague Cedex (France)

    2006-07-01

    The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

  2. Burn-up measurements coupling gamma spectrometry and neutron measurement

    International Nuclear Information System (INIS)

    Toubon, H.; Pin, P.; Lebrun, A.; Oriol, L.; Saurel, N.; Gain, T.

    2006-01-01

    The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)

  3. New Burnup Calculation System for Fusion-Fission Hybrid System

    International Nuclear Information System (INIS)

    Isao Murata; Shoichi Shido; Masayuki Matsunaka; Keitaro Kondo; Hiroyuki Miyamaru

    2006-01-01

    Investigation of nuclear waste incineration has positively been carried out worldwide from the standpoint of environmental issues. Some candidates such as ADS, FBR are under discussion for possible incineration technology. Fusion reactor is one of such technologies, because it supplies a neutron-rich and volumetric irradiation field, and in addition the energy is higher than nuclear reactor. However, it is still hard to realize fusion reactor right now, as well known. An idea of combination of fusion and fission concepts, so-called fusion-fission hybrid system, was thus proposed for the nuclear waste incineration. Even for a relatively lower plasma condition, neutrons can be well multiplied by fission in the nuclear fuel, tritium is thus bred so as to attain its self-sufficiency, enough energy multiplication is then expected and moreover nuclear waste incineration is possible. In the present study, to realize it as soon as possible with the presently proven technology, i.e., using ITER model with the achieved plasma condition of JT60 in JAEA, Japan, a new calculation system for fusion-fission hybrid reactor including transport by MCNP and burnup by ORIGEN has been developed for the precise prediction of the neutronics performance. The author's group already has such a calculation system developed by them. But it had a problem that the cross section libraries in ORIGEN did not have a cross section library, which is suitable specifically for fusion-fission hybrid reactors. So far, those for FBR were approximately used instead in the analysis. In the present study, exact derivation of the collapsed cross section for ORIGEN has been investigated, which means it is directly evaluated from calculated track length by MCNP and point-wise nuclear data in the evaluated nuclear data file like JENDL-3.3. The system realizes several-cycle calculation one time, each of which consists of MCNP criticality calculation, MCNP fixed source calculation with a 3-dimensional precise

  4. The burn-up credit physics and the 40. Minerve anniversary; La physique du credit Burn-Up et le 40. anniversaire de Minerve

    Energy Technology Data Exchange (ETDEWEB)

    Santamarina, A [CEA/Cadarache, Departement d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Toubon, H [Cogema, 78 - Velizy Villacoublay (France); Trakas, C [FRAMATOME, 92 - Paris La Defense (France); and others

    2000-03-21

    The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)

  5. Corrections to the 148Nd method of evaluation of burnup for the PIE samples from Mihama-3 and Genkai-1 reactors

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki

    2006-01-01

    The value of the burnup is one of the most important parameters of samples taken by post-irradiation examination (PIE). Generally, it is evaluated by the Neodymium-148 method. Precise evaluation of the burnup value requires: (1) an effective fission yield of 148 Nd; (2) neutron capture reactions of 147 Nd and 148 Nd; (3) a conversion factor from fissions per initial heavy metal to the burnup unit GWd/t. In this study, the burnup values of the PIE data from Mihama-3 and Genkai-1 PWRs, which were taken by the Japan Atomic Energy Research Institute, were re-evaluated using more accurate corrections for each of these three items. The PIE data were then re-analyzed using SWAT and SWAT2 code systems with JENDL-3.3 library. The re-evaluation of the effective fission yield of 148 Nd has an effect of 1.5-2.0% on burnup values. Considering the neutron capture reactions of 147 Nd and 148 Nd removes dependence of C/E values of 148 Nd on the burnup value. The conversion factor from FIMA(%) to GWd/t changes according to the burnup value. Its effect on the burnup evaluation is small for samples having burnup of larger than 30 GWd/t. The analyses using the corrected burnup values showed that the calculated 148 Nd concentrations and the PIE data is approximately 1%, whereas this was 3-5% in prior analyses. This analysis indicates that the burnup values of samples from Mihama-3 and Genkai-1 PWRs should be corrected by 2-3%. The effect of re-evaluation of the burnup value on the neutron multiplication factor is an approximately 0.6% change in PIE samples having the burnup of larger than 30 GWd/t. Finally, comparison between calculation results using a single pin-cell model and an assembly model is carried out. Because the results agreed with each other within a few percent, we concluded that the single pin-cell model is suitable for the analysis of PIE samples and that the underestimation of plutonium isotopes, which occurred in the previous analyses, does not result from a geometry

  6. Corrections to the {sup 148}Nd method of evaluation of burnup for the PIE samples from Mihama-3 and Genkai-1 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya [Fuel Cycle Facility Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)]. E-mail: suyama.kenya@jaea.go.jp; Mochizuki, Hiroki [Japan Research Institute, Limited, 16 Ichiban-cho, Chiyoda-ku, Tokyo 102-0082 (Japan)

    2006-03-15

    The value of the burnup is one of the most important parameters of samples taken by post-irradiation examination (PIE). Generally, it is evaluated by the Neodymium-148 method. Precise evaluation of the burnup value requires: (1) an effective fission yield of {sup 148}Nd; (2) neutron capture reactions of {sup 147}Nd and {sup 148}Nd; (3) a conversion factor from fissions per initial heavy metal to the burnup unit GWd/t. In this study, the burnup values of the PIE data from Mihama-3 and Genkai-1 PWRs, which were taken by the Japan Atomic Energy Research Institute, were re-evaluated using more accurate corrections for each of these three items. The PIE data were then re-analyzed using SWAT and SWAT2 code systems with JENDL-3.3 library. The re-evaluation of the effective fission yield of {sup 148}Nd has an effect of 1.5-2.0% on burnup values. Considering the neutron capture reactions of {sup 147}Nd and {sup 148}Nd removes dependence of C/E values of {sup 148}Nd on the burnup value. The conversion factor from FIMA(%) to GWd/t changes according to the burnup value. Its effect on the burnup evaluation is small for samples having burnup of larger than 30 GWd/t. The analyses using the corrected burnup values showed that the calculated {sup 148}Nd concentrations and the PIE data is approximately 1%, whereas this was 3-5% in prior analyses. This analysis indicates that the burnup values of samples from Mihama-3 and Genkai-1 PWRs should be corrected by 2-3%. The effect of re-evaluation of the burnup value on the neutron multiplication factor is an approximately 0.6% change in PIE samples having the burnup of larger than 30 GWd/t. Finally, comparison between calculation results using a single pin-cell model and an assembly model is carried out. Because the results agreed with each other within a few percent, we concluded that the single pin-cell model is suitable for the analysis of PIE samples and that the underestimation of plutonium isotopes, which occurred in the previous

  7. Automated generation of burnup chain for reactor analysis applications

    International Nuclear Information System (INIS)

    Tran Viet Phu; Tran Hoai Nam; Akio Yamamoto; Tomohiro Endo

    2015-01-01

    This paper presents the development of an automated generation of a new burnup chain for reactor analysis applications. The JENDL FP Decay Data File 2011 and Fission Yields Data File 2011 were used as the data sources. The nuclides in the new chain are determined by restrictions of the half-life and cumulative yield of fission products or from a given list. Then, decay modes, branching ratios and fission yields are recalculated taking into account intermediate reactions. The new burnup chain is output according to the format for the SRAC code system. Verification was performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Further development and applications are being planned with the burnup chain code. (author)

  8. Probabilistic assessment of dry transport with burnup credit

    International Nuclear Information System (INIS)

    Lake, W.H.

    2003-01-01

    The general concept of probabilistic analysis and its application to the use of burnup credit in spent fuel transport is explored. Discussion of the probabilistic analysis method is presented. The concepts of risk and its perception are introduced, and models are suggested for performing probability and risk estimates. The general probabilistic models are used for evaluating the application of burnup credit for dry spent nuclear fuel transport. Two basic cases are considered. The first addresses the question of the relative likelihood of exceeding an established criticality safety limit with and without burnup credit. The second examines the effect of using burnup credit on the overall risk for dry spent fuel transport. Using reasoned arguments and related failure probability and consequence data analysis is performed to estimate the risks of using burnup credit for dry transport of spent nuclear fuel. (author)

  9. Burnup credit applications in a high-capacity truck cask

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1992-09-01

    General Atomics (GA) has designed two legal weight truck (LWT) casks, the GA-4 and GA-9, to carry four pressurized-water-reactor (PWR) and nine boiling-water-reactor (BWR) fuel assemblies, respectively. GA plans to submit applications for certification to the US Nuclear Regulatory Commission (NRC) for the two casks in mid-1993. GA will include burnup credit analysis in the Safety Analysis Report for Packaging (SARP) for the GA-4 Cask. By including burnup credit in the criticality safety analysis for PWR fuels with initial enrichments above 3% U-235, public and occupation risks are reduced and cost savings are realized. The GA approach to burnup credit analysis incorporates the information produced in the US Department of Energy Burnup Credit Program. This paper describes the application of burnup credit to the criticality control design of the GA-4 Cask

  10. Modelling of high burnup structure in UO2 fuel with the RTOP code

    International Nuclear Information System (INIS)

    Likhanskii, V.; Zborovskii, V.; Evdokimov, I.; Kanyukova, V.; Sorokin, A.

    2008-01-01

    The present work deals with self-consistent physical approach aimed to derive the criterion of fuel restructuring avoiding correlations. The approach is based on study of large over pressurized bubbles formation on dislocations, at grain boundaries and in grain volume. At first, stage of formation of bubbles non-destroyable by fission fragments is examined using consistent modelling of point defects and fission gas behavior near dislocation and in grain volume. Then, evolution of formed large non-destroyable bubbles is considered using results of the previous step as initial values. Finally, condition of dislocation loops punching by sufficiently large over pressurized bubbles is regarded as the criterion of fuel restructuring onset. In the present work consideration of large over pressurized bubbles evolution is applied to modelling of the restructuring threshold depending on temperature, burnup and grain size. Effect of grain size predicted by the model is in qualitative agreement with experimental observations. Restructuring threshold criterion as an analytical function of local burnup and fuel temperature is derived and compared with HBRP project data. To predict rim-layer width formation depending on fuel burnup and irradiation conditions the model is implemented into the mechanistic fuel performance code RTOP. Calculated dependencies give upper estimate for the width of restructured region. Calculations show that one needs to consider temperature distribution within pellet which depends on irradiation history in order to model rim-structure formation

  11. Benefits of actinide-only burnup credit for shutdown PWRs

    International Nuclear Information System (INIS)

    Lancaster, D.; Fuentes, E.; Kang, C.; Rivard, D.

    1998-02-01

    Owners of PWRs that are shutdown prior to resolution of interim storage or permanent disposal issues have to make difficult decisions on what to do with their spent fuel. Maine Yankee is currently evaluating multiple options for spent fuel storage. Their spent fuel pool has 1,434 assemblies. In order to evaluate the value to a utility of actinide-only burnup credit, analysis of the number of canisters required with and without burnup credit was made. In order to perform the analysis, loading curves were developed for the Holtec Hi-Star 100/MPC-32. The MPC-32 is hoped to be representative of future burnup credit designs from many vendors. The loading curves were generated using the actinide-only burnup credit currently under NRC review. The canister was analyzed for full loading (32 assemblies) and with partial loadings of 30 and 28 assemblies. If no burnup credit is used the maximum capacity was assumed to be 24 assemblies. this reduced capacity is due to the space required for flux traps which are needed to sufficiently reduce the canister reactivity for the fresh fuel assumption. Without burnup credit the 1,343 assemblies would require 60 canisters. If all the fuel could be loaded into the 32 assembly canisters only 45 canisters would be required. Although the actinide-only burnup credit approach is very conservative, the total number of canisters required is only 47 which is only two short of the minimum possible number of canisters. The utility is expected to buy the canister and the storage overpack. A reasonable cost estimate for the canister plus overpack is $500,000. Actinide-only burnup credit would save 13 canisters and overpacks which is a savings of about $6.5 million. This savings is somewhat reduced since burnup credit requires a verification measurement of burnup. The measurement costs for these assemblies can be estimated as about $1 million. The net savings would be $5.5 million

  12. Monte Carlo burnup codes acceleration using the correlated sampling method

    International Nuclear Information System (INIS)

    Dieudonne, C.

    2013-01-01

    For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes to simulate the neutron transport to deterministic methods, which handle the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3-dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the Monte Carlo solver called at each time step. In this document we present an original methodology to avoid the repetitive and time-expensive Monte Carlo simulations, and to replace them by perturbation calculations: indeed the different burnup steps may be seen as perturbations of the isotopic concentration of an initial Monte Carlo simulation. In a first time we will present this method, and provide details on the perturbative technique used, namely the correlated sampling. In a second time we develop a theoretical model to study the features of the correlated sampling method to understand its effects on depletion calculations. In a third time the implementation of this method in the TRIPOLI-4 code will be discussed, as well as the precise calculation scheme used to bring important speed-up of the depletion calculation. We will begin to validate and optimize the perturbed depletion scheme with the calculation of a REP-like fuel cell depletion. Then this technique will be used to calculate the depletion of a REP-like assembly, studied at beginning of its cycle. After having validated the method with a reference calculation we will show that it can speed-up by nearly an order of magnitude standard Monte-Carlo depletion codes. (author) [fr

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

    International Nuclear Information System (INIS)

    Lancaster, D.; Machiels, A.

    2012-01-01

    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 2 critical experiments to determine the criticality safety limits on the neutron multiplication factor, k 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Las Vegas, NV (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

  16. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  17. The commercial impact of burnup increase

    International Nuclear Information System (INIS)

    Fenzlein, C.; Schricker, W.

    2002-01-01

    Deregulation has a dramatic effect on competition in the electricity markets. This will lead to a continued pressure on the prices in virtually all areas of the nuclear fuel cycle and will encourage further optimization, technical and technological progress and innovations with respect to further cost reductions of power production. The permission of direct disposal, in Germany legally granted in 1994 as an alternative to the reprocessing path, made possible cost savings and has consequently resulted in a decline of reprocessing prices. In addition, suppliers as well as operators are making considerable efforts to reduce the disposal costs fraction by optimizing disposal technologies and concepts. The increase of discharge has essentially contributed to the reduction the disposal cost fraction. Compared to former scenarios, the economic potential of burn-up increase is decreasing

  18. Measurement of burnup in FBR MOX fuel irradiated to high burnup

    International Nuclear Information System (INIS)

    Koyama, Shin-ichi; Osaka, Masahiko; Sekine, Takashi; Morozumi, Katsufumi; Namekawa, Takashi; Itoh, Masahiko

    2003-01-01

    The burnup of fuel pins in the subassemblies irradiated at the range from 0.003 to 13.28% FIMA in the JOYO MK-II core were measured by the isotope dilution analysis. For the measurement, 75 and 51 specimens were taken from the fuel pins of driver fuel and irradiation test subassemblies, respectively. The data of burnup could be obtained within an experimental error of 4%, and were compared with the ones calculated by 3-dimensional neutron diffusion codes MAGI and ESPRIT-J, which are used for JOYO core management system. Both data of burnup almost agree with each other within an error of 5%. For the fuel pins loaded at the outer region of the subassembly in the 4th row, which was adjacent to reflectors, however, some of the calculation results were 15% less at most than the measured values. It is suggested from the calculation by a Monte Carlo code MCNP-4A that this difference between the calculated and the measured data attribute from the softening of neutron flux in the region adjacent to the reflector. (author)

  19. High burnup models in computer code fair

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [Bhabha Atomic Research Centre, Bombay (India)

    1997-08-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.

  20. High burnup models in computer code fair

    International Nuclear Information System (INIS)

    Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.

    1997-01-01

    An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs

  1. Effect of burnup on the response of stainless steel-clad mixed-oxide fuels to simulated thermal transients

    International Nuclear Information System (INIS)

    Fenske, G.R.; Badyopadhyay, G.

    1981-01-01

    Direct electrical heating experiments were performed on irradiated fuel to study the fuel and cladding response as a function of burnup during a slow thermal transient. The results indicated that the nature and extent of the fuel and cladding behavior depended on the quantity of fission gas retained in the fuel. Fission-gas-driven fuel ejection occurred as the molten cladding flowed down the stack exposing bare, radially unrestrained fuel. The fuel dispersion occurred in the absence of molten fuel and the amount of fuel ejected increased with increasing burnup. 31 refs

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

    International Nuclear Information System (INIS)

    BSC

    2004-01-01

    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

  3. Light a CANDLE. An innovative burnup strategy of nuclear reactors

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi

    2005-11-01

    CANDLE is a new burnup strategy for nuclear reactors, which stands for Constant Axial Shape of Neutron Flux, Nuclide Densities and Power Shape During Life of Energy Production. When this candle-like burnup strategy is adopted, although the fuel is fixed in a reactor core, the burning region moves, at a speed proportionate to the power output, along the direction of the core axis without changing the spatial distribution of the number density of the nuclides, neutron flux, and power density. Excess reactivity is not necessary for burnup and the shape of the power distribution and core characteristics do not change with the progress of burnup. It is not necessary to use control rods for the control of the burnup. This booklet described the concept of the CANDLE burnup strategy with basic explanations of excess neutrons and its specific application to a high-temperature gas-cooled reactor and a fast reactor with excellent neutron economy. Supplementary issues concerning the initial core and high burnup were also referred. (T. Tanaka)

  4. BNFL assessment of methods of attaining high burnup MOX fuel

    International Nuclear Information System (INIS)

    Brown, C.; Hesketh, K.W.; Palmer, I.D.

    1998-01-01

    It is clear that in order to maintain competitiveness with UO 2 fuel, the burnups achievable in MOX fuel must be enhanced beyond the levels attainable today. There are two aspects which require attention when studying methods of increased burnups - cladding integrity and fuel performance. Current irradiation experience indicates that one of the main performance issues for MOX fuel is fission gas retention. MOX, with its lower thermal conductivity, runs at higher temperatures than UO 2 fuel; this can result in enhanced fission gas release. This paper explores methods of effectively reducing gas release and thereby improving MOX burnup potential. (author)

  5. A guide to introducing burnup credit, preliminary version (English translation)

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu

    2017-06-01

    There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee. (author)

  6. Burnup credit demands for spent fuel management in Ukraine

    International Nuclear Information System (INIS)

    Medun, V.

    2001-01-01

    In fact, till now, burnup credit has not be applied in Ukrainian nuclear power for spent fuel management systems (storage and transport). However, application of advanced fuel at VVER reactors, arising spent fuel amounts, represent burnup credit as an important resource to decrease spent fuel management costs. The paper describes spent fuel management status in Ukraine from viewpoint of subcriticality assurance under spent fuel storage and transport. It also considers: 1. Regulation basis concerning subcriticality assurance, 2. Basic spent fuel and transport casks characteristics, 3. Possibilities and demands for burnup credit application at spent fuel management systems in Ukraine. (author)

  7. Burnup calculation methodology in the serpent 2 Monte Carlo code

    International Nuclear Information System (INIS)

    Leppaenen, J.; Isotalo, A.

    2012-01-01

    This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)

  8. A microcomputer program for coupled cycle burnup calculations

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Downar, T.J.; Taylor, E.L.

    1986-01-01

    A program, designated BRACC (Burnup, Reactivity, And Cycle Coupling), has been developed for fuel management scoping calculations, and coded in the BASIC language in an interactive format for use with microcomputers. BRACC estimates batch and cycle burnups for sequential reloads for a variety of initial core conditions, and permits the user to specify either reload batch properties (enrichment, burnable poison reactivity) or the target cycle burnup. Most important fuel management tactics (out-in or low-leakage loading, coastdown, variation in number of assemblies charged) can be simulated

  9. Safety aspects related to burnup increase and mixed oxide fuel

    International Nuclear Information System (INIS)

    Thomas, W.

    1992-01-01

    The dominant factor presently limiting the fuel burnup is the response of the cladding hulls. To maintain the excellent record of very low fuel failure rates for increased burnups further technical development is underway and necessary. In the nuclear fuel cycle increased burnups lead to a remarkable reduction of spent fuel arisings and corresponding economic savings. Thermal recycling of plutonium presently provides an opportunity to reduce the rising accumulation of plutunium in a situation where there is no demand for this fissile material in Fast Breeder Reactors. (orig.) [de

  10. Two dimensional burn-up calculation of TRIGA core

    International Nuclear Information System (INIS)

    Persic, A.; Ravnik, M.; Slavic, S.

    1996-01-01

    TRIGLAV is a new computer program for burn-up calculation of mixed core of research reactors. The code is based on diffusion model in two dimensions and iterative procedure is applied for its solution. The material data used in the model are calculated with the transport program WIMS. In regard to fission density distribution and energy produced by the reactor the burn-up increment of fuel elements is determined. In this paper the calculation model of diffusion constants and burn-up calculation are described and some results of calculations for TRIGA MARK II reactor are presented. (author)

  11. Measurement and interpretation of triton burnup in Jet deuterium plasmas

    International Nuclear Information System (INIS)

    Jarvis, O.N.; Kallne, J.; Sadler, G.; van Belle, P.; Gorini, G.; Conroy, S.; Verschuur, K.

    1989-01-01

    The confinement and slowing down of fast tritons in JET deuterium plasmas is investigated. The ratio of 14 MeV and 2.5 MeV neutron production rates is measured. This ratio is equal to the fraction of tritons which burnup. The 2.5 MeV neutron emission is obtained from a set of fission chambers for which the calibration uncertainty is about 10%. The absolute calibration of the activation technique is calculated. The comparison between experimental and theoretical burnup ratios, for JET 1987 data, is shown. The range of conditions over which measurements of triton burnup fraction were obtained, is illustrated

  12. Effectiveness of Mindfulness-Based Group Therapy Compared to the Usual Opioid Dependence Treatment.

    Science.gov (United States)

    Imani, Saeed; Atef Vahid, Mohammad Kazem; Gharraee, Banafsheh; Noroozi, Alireza; Habibi, Mojtaba; Bowen, Sarah

    2015-06-01

    This study investigated the effectiveness of mindfulness-based group therapy (MBGT) compared to the usual opioid dependence treatment (TAU).Thirty outpatients meeting the DSM-IV-TR criteria for opioid dependence from Iranian National Center for Addiction Studies (INCAS) were randomly assigned into experimental (Mindfulness-Based Group Therapy) and control groups (the Usual Treatment).The experimental group undertook eight weeks of intervention, but the control group received the usual treatment according to the INCAS program. The Five Factor Mindfulness Questionnaire (FFMQ) and the Addiction Sevier Index (ASI) were administered at pre-treatment and post-treatment assessment periods. Thirteen patients from the experimental group and 15 from the control group completed post-test assessments. The results of MANCOVA revealed an increase in mean scores in observing, describing, acting with awareness, non-judging, non-reacting, and decrease in mean scores of alcohol and opium in MBGT patient group. The effectiveness of MBGT, compared to the usual treatment, was discussed in this paper as a selective protocol in the health care setting for substance use disorders.

  13. Effectiveness of Mindfulness-Based Group Therapy Compared to the Usual Opioid Dependence Treatment

    Directory of Open Access Journals (Sweden)

    Saeed Imani

    2015-11-01

    Full Text Available  Objective: This study investigated the effectiveness of mindfulness-based group therapy (MBGT compared to the usual opioid dependence treatment (TAU.Thirty outpatients meeting the DSM-IV-TR criteria for opioid dependence from Iranian National Center for Addiction Studies (INCAS were randomly assigned into experimental (Mindfulness-Based Group Therapy and control groups (the Usual Treatment.The experimental group undertook eight weeks of intervention, but the control group received the usual treatment according to the INCAS program.  Methods:The Five Factor Mindfulness Questionnaire (FFMQ and the Addiction Sevier Index (ASI were administered at pre-treatment and post-treatment assessment periods. Thirteen patients from the experimental group and 15 from the control group completed post-test assessments. Results:The results of MANCOVA revealed an increase in mean scores in observing, describing, acting with awareness, non-judging, non-reacting, and decrease in mean scores of alcohol and opium in MBGT patient group. Conclusion:The effectiveness of MBGT, compared to the usual treatment, was discussed in this paper as a selective protocol in the health care setting for substance use disorders.

  14. Micrographic study on distribution of fission products in high burn-up metallic alloy fuel

    International Nuclear Information System (INIS)

    Kolay, S.; Basu, M.; Das, D.

    2012-01-01

    One of the important mandates in the three-stage nuclear power generation programme of India is to utilize uranium-plutonium based alloy fuels in enabling shorter doubling time for breeding of the fissile isotopes ( 239 Pu and 233 U ) to be used in thorium based driver fuel in the third stage. Reported information shows the successful performance of alloy fuel with somewhat porous matrix in achieving 10-15 atom% burnup. The porosity and microstructure of these alloys are strongly dependent on their composition and phases present. Porosity also influences the extent of fuel swelling and gas release. So to assess fuel performance and fuel integrity under high burn-up condition it is essential to have knowledge about the new phases formed and their redistribution that occurs as a result of inter-diffusion and temperature gradient. This study addresses these issues taking the base alloy U-10 wt %Zr

  15. Conductance of Conjugated Molecular Wires: Length Dependence, Anchoring Groups, and Band Alignment

    DEFF Research Database (Denmark)

    Peng, Guowen; Strange, Mikkel; Thygesen, Kristian Sommer

    2009-01-01

    , is not solely determined by the intrinsic band gap of the molecular wire but also depends on the anchoring group. This is because the alignment of the metal Fermi level with respect to the molecular levels is controlled by charge transfer and interface dipoles which in turn are determined by the local chemistry...

  16. Calculation Of Fuel Burnup And Radionuclide Inventory In The Syrian Miniature Neutron Source Reactor Using The GETERA Code

    International Nuclear Information System (INIS)

    Khattab, K.; Dawahra, S.

    2011-01-01

    Calculations of the fuel burnup and radionuclide inventory in the Syrian Miniature Neutron Source Reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burnup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission product and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer and has a bigger library of isotopes and more accurate. (author)

  17. 3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

    International Nuclear Information System (INIS)

    Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

    2013-01-01

    Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

  18. The use of burnup credit for spent fuel cask design

    International Nuclear Information System (INIS)

    Lake, W.H.

    1993-01-01

    A new generation of high capacity spent fuel transport casks is being developed by the U.S. Department of Energy (DOE) as part of the Federal Waste Management System (FWMS). Burnup credit, which recognizes the reduced reactivity of spent fuel is being used for these casks. Two cask designs being developed for DOE by Babcock and Wilcox and General Atomics use burnup credit. The cask designs must be certified by the U.S. Nuclear Regulatory Commission (NRC) if they are to be used in the FWMS. Certification of these casks by the NRC would not require any change in the NRC's transport regulations, and would be consistent with past practices. Furthermore, use of burnup credit casks appears to be consistent with current International Atomic Energy Agency (IAEA) rules and regulations. To support NRC certification, DOE has identified the technical issues related to burnup credit, and embarked on a development program to resolve them. (J.P.N.)

  19. Burn-up measurement in the HTR-module-reactor

    International Nuclear Information System (INIS)

    Gerhards, E.

    1993-05-01

    The burn-up status of spherical HTR-fuel elements is determined by a γ-spectrometric analysis of Cs-137 activity. The γ-spectrum recorded by a semiconductor detector up to now is analyzed by complex mathematical and time-consuming methods. For the operation of the HTR-Module-Reactor, however, a fast evaluation of the burn-up status is necessary. It is shown that this can be ensured by a comparison between the measured spectra and simulation results. Using the computer-program HTROGEN and the program system SPECCALC especially developed for this problem the γ-spectra are evaluated as a function of the burn-up status. The method is applied to results available from the operation of the AVR-reactor. The burn-up status determined with different methods corresponds very well within the limits of accuracy. (orig.)

  20. Parametric neutronic analyses related to burnup credit cask design

    International Nuclear Information System (INIS)

    Parks, C.V.

    1989-01-01

    The consideration of spent fuel histories (burnup credit) in the design of spent fuel shipping casks will result in cost savings and public risk benefits in the overall fuel transportation system. The purpose of this paper is to describe the depletion and criticality analyses performed in conjunction with and supplemental to the referenced analysis. Specifically, the objectives are to indicate trends in spent fuel isotopic composition with burnup and decay time; provide spent fuel pin lattice values as a function of burnup, decay time, and initial enrichment; demonstrate the variation of k eff for infinite arrays of spent fuel assemblies separated by generic cask basket designs (borated and unborated) of varying thicknesses; and verify the potential cask reactivity margin available with burnup credit via analysis with generic cask models

  1. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

    Proselkov, V.N.; Scheglov, A.S.; Smirnov, A.V.; Smirnov, V.P.

    2001-01-01

    Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

  2. An optimal model for fuel burnup in nuclear reactors

    International Nuclear Information System (INIS)

    Anton, V.

    1979-05-01

    An approach to minimize the number of the burnup equations taking into account the introduction of an appropriate number of fission products is given. The corresponding number of fission pseudo-products is defined. (author)

  3. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Hyoudou, Hideaki; Tatsumi, Masahiro

    2010-02-01

    Improvement on prediction accuracy for neutronics property of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified constants library as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores, however, improvement of not only static properties but also burnup properties is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup properties using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous research, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for production systems. In the present study, we implemented functions for cell calculations and burnup calculations. With this, whole steps in analysis can be carried out with only this system. In addition, we modified the specification of user input to improve the convenience of this system. Since implementations being done so

  4. A guide introducing burnup credit, preliminary version. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    It is examined to take burnup credit into account for criticality safety control of facility treating spent fuel. This work is a collection of current technical status of predicting isotopic composition and criticality of spent fuel, points to be specially considered for safety evaluation, and current status of legal affairs for the purpose of applying burnup credit to the criticality safety evaluation of the facility treating spent fuel in Japan. (author)

  5. A burn-up module coupling to an AMPX system

    International Nuclear Information System (INIS)

    Salvatore Duque, M.; Gomez, S.E.; Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.

    1990-01-01

    The Reactors and Neutrons Division of the Bariloche Atomic Center uses the AMPX system for the study of high conversion reactors (HCR). Such system allows to make neutronic calculations from the nuclear data library (ENDF/B-IV). The Nuclear Engineering career of the Balseiro Institute developed and implemented a burn-up module at a μ-cell level (BUM: Burn-up Module) which agrees with the requirement to be coupled to the AMPX system. (Author) [es

  6. Application of dynamic pseudo fission products and actinides for accurate burnup calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.; Kloosterman, J.L.

    1996-09-01

    The introduction of pseudo fission products for accurate fine-group spectrum calculations during burnup is discussed. The calculation of the density of the pseudo nuclides is done before each spectrum calculation from the actual densities and their cross sections of all nuclides to be lumped into a pseudo fission product. As there are also many actinides formed in the fuel during its life cycle, a pseudo actinide with fission cross section is also introduced. From a realistic burnup calculation it is demonstrated that only a few fission products and actinides need to be included explicitly in a spectrum calculation. All other fission products and actinides can be accurately represented in the pseudo nuclides. (author)

  7. A comparative study to investigate burnup in research reactor fuel using two independent experimental methods

    International Nuclear Information System (INIS)

    Iqbal, M.; Mehmood, T.; Ayazuddin, S.K.; Salahuddin, A.; Pervez, S.

    2001-01-01

    Two independent experimental methods have been used for the comparative study of fuel burnup measurement in low enriched uranium, plate type research reactor. In the first method a gamma ray activity ratio method was employed. An experimental setup was established for gamma ray scanning using prior calibrated high purity germanium detector. The computer software KORIGEN gave the theoretical support. In the second method reactivity difference technique was used. At the same location in the same core configuration the fresh and burned fuel element's reactivity worth was estimated. For theoretical estimated curve, group cross-sections were generated using computer code WIMS-D/4, and three dimensional modeling was made by computer code CITATION. The measured burnup of different fuel elements using these methods were found to be in good agreement

  8. The burn-up credit physics and the 40. Minerve anniversary

    International Nuclear Information System (INIS)

    Santamarina, A.; Toubon, H.; Trakas, C.

    2000-01-01

    The technical meeting organized by the SFEN on the burn-up credit (CBU) physics, took place the 23 november 1999 at Cadarache. the first presentation dealt with the economic interest and the neutronic problems of the CBU. Then two papers presented how taking into account the CBU in the industry in matter of transport, storage in pool, reprocessing and criticality calculation (MCNP4/Apollo2-F benchmark). An experimental method for the reactivity measurement through oscillations in the Minerve reactor, has been presented with an analysis of the possible errors. The future research program OSMOSE, taking into account the minor actinides in the CBU, was also developed. The last paper presented the national and international research programs in the CBU domain, in particular experiments realized in CEA/Valduc and the OECD Burn-up Criticality Benchmark Group activities. (A.L.B.)

  9. Development of a Burnup Module DECBURN Based on the Krylov Subspace Method

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J. Y.; Kim, K. S.; Shim, H. J.; Song, J. S

    2008-05-15

    This report is to develop a burnup module DECBURN that is essential for the reactor analysis and the assembly homogenization codes to trace the fuel composition change during the core burnup. The developed burnup module solves the burnup equation by the matrix exponential method based on the Krylov Subspace method. The final solution of the matrix exponential is obtained by the matrix scaling and squaring method. To develop DECBURN module, this report includes the followings as: (1) Krylov Subspace Method for Burnup Equation, (2) Manufacturing of the DECBURN module, (3) Library Structure Setup and Library Manufacturing, (4) Examination of the DECBURN module, (5) Implementation to the DeCART code and Verification. DECBURN library includes the decay constants, one-group cross section and the fission yields. Examination of the DECBURN module is performed by manufacturing a driver program, and the results of the DECBURN module is compared with those of the ORIGEN program. Also, the implemented DECBURN module to the DeCART code is applied to the LWR depletion benchmark and a OPR-1000 pin cell problem, and the solutions are compared with the HELIOS code to verify the computational soundness and accuracy. In this process, the criticality calculation method and the predictor-corrector scheme are introduced to the DeCART code for a function of the homogenization code. The examination by a driver program shows that the DECBURN module produces exactly the same solution with the ORIGEN program. DeCART code that equips the DECBURN module produces a compatible solution to the other codes for the LWR depletion benchmark. Also the multiplication factors of the DeCART code for the OPR-1000 pin cell problem agree to the HELIOS code within 100 pcm over the whole burnup steps. The multiplication factors with the criticality calculation are also compatible with the HELIOS code. These results mean that the developed DECBURN module works soundly and produces an accurate solution

  10. Competing failure analysis in phased-mission systems with multiple functional dependence groups

    International Nuclear Information System (INIS)

    Wang, Chaonan; Xing, Liudong; Peng, Rui; Pan, Zhusheng

    2017-01-01

    A phased-mission system (PMS) involves multiple, consecutive, non-overlapping phases of operation. The system structure function and component failure behavior in a PMS can change from phase to phase, posing big challenges to the system reliability analysis. Further complicating the problem is the functional dependence (FDEP) behavior where the failure of certain component(s) causes other component(s) to become unusable or inaccessible or isolated. Previous studies have shown that FDEP can cause competitions between failure propagation and failure isolation in the time domain. While such competing failure effects have been well addressed in single-phase systems, only little work has focused on PMSs with a restrictive assumption that a single FDEP group exists in one phase of the mission. Many practical systems (e.g., computer systems and networks), however may involve multiple FDEP groups during the mission. Moreover, different FDEP groups can be dependent due to sharing some common components; they may appear in a single phase or multiple phases. This paper makes new contributions by modeling and analyzing reliability of PMSs subject to multiple FDEP groups through a Markov chain-based methodology. Propagated failures with both global and selective effects are considered. Four case studies are presented to demonstrate application of the proposed method. - Highlights: • Reliability of phased-mission systems subject to competing failure propagation and isolation effects is modeled. • Multiple independent or dependent functional dependence groups are considered. • Propagated failures with global effects and selective effects are studied. • Four case studies demonstrate generality and application of the proposed Markov-based method.

  11. Application of Candle burnup to small fast reactor

    International Nuclear Information System (INIS)

    Sekimoto, H.; Satoshi, T.

    2004-01-01

    A new reactor burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. An equilibrium state was obtained for a large fast reactor (core radius is 2 m and reflector thickness is 0.5 m) successfully by using a newly developed direct analysis code. However, it is difficult to apply this burnup strategy to small reactors, since its neutron leakage becomes large and neutron economy becomes worse. Fuel enrichment should be increased in order to sustain the criticality. However, higher enrichment of fresh fuel makes the CANDLE burnup difficult. We try to find some small reactor designs, which can realize the CANDLE burnup. We have successfully find a design, which is not the CANDLE burnup in the strict meaning, but satisfies qualitatively its characteristics mentioned at the top of this abstract. In the final paper, the general description of CANDLE burnup and some results on the obtained small fast reactor design are presented.(author)

  12. Monte Carlo burnup simulation of the TAKAHAMA-3 benchmark experiment

    International Nuclear Information System (INIS)

    Dalle, Hugo M.

    2009-01-01

    High burnup PWR fuel is currently being studied at CDTN/CNEN-MG. Monte Carlo burnup code system MONTEBURNS is used to characterize the neutronic behavior of the fuel. In order to validate the code system and calculation methodology to be used in this study the Japanese Takahama-3 Benchmark was chosen, as it is the single burnup benchmark experimental data set freely available that partially reproduces the conditions of the fuel under evaluation. The burnup of the three PWR fuel rods of the Takahama-3 burnup benchmark was calculated by MONTEBURNS using the simplest infinite fuel pin cell model and also a more complex representation of an infinite heterogeneous fuel pin cells lattice. Calculations results for the mass of most isotopes of Uranium, Neptunium, Plutonium, Americium, Curium and some fission products, commonly used as burnup monitors, were compared with the Post Irradiation Examinations (PIE) values for all the three fuel rods. Results have shown some sensitivity to the MCNP neutron cross-section data libraries, particularly affected by the temperature in which the evaluated nuclear data files were processed. (author)

  13. Modelling the high burnup UO2 structure in LWR fuel

    International Nuclear Information System (INIS)

    Lassmann, K.; Walker, C.T.; Laar, J. van de; Lindstroem, F.

    1995-01-01

    The concept of a burnup threshold for the formation of the high burnup UO 2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO 2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60-75 GW d/t U. The lower value is considered to be the threshold for the onset of the HBS and the higher value the threshold for the fully developed HBS. Xenon depletion in the transition zone and the fully developed HBS can be described by a simple model. At local burnups above 120 GW d/t U the xenon generated is in equilibrium with the xenon lost to the fission gas pores and the concentration does not fall below 0.25 wt%. The TRANSURANUS burnup model TUBRNP predicts reasonably well the penetration of the HBS and the associated xenon depletion up to a cross section average burnup of approximately 70 GW d/t U. (orig.)

  14. EVOLUT - a computer program for fast burnup evaluation

    International Nuclear Information System (INIS)

    Craciunescu, T.; Dobrin, R.; Stamatescu, L.; Alexa, A.

    1999-01-01

    EVOLUT is a computer program for burnup evaluation. The input data consist on the one hand of axial and radial gamma-scanning profiles (for the experimental evaluation of the number of nuclei of a fission product - the burnup monitor - at the end of irradiation) and on the other hand of the history of irradiation (the time length and values proportional to the neutron flux for each step of irradiation). Using the equation of evolution of the burnup monitor the flux values are iteratively adjusted, by a multiplier factor, until the calculated number of nuclei is equal to the experimental one. The flux values are used in the equation of evolution of the fissile and fertile nuclei to determine the fission number and consequently the burnup. EVOLUT was successfully used in the analysis of several hundreds of CANDU and TRIGA-type fuel rods. We appreciate that EVOLUT is a useful tool in the burnup evaluation based on gamma spectrometry measurements. EVOLUT can be used on an usual AT computer and in this case the results are obtained in a few minutes. It has an original and user-friendly graphical interface and it provides also output in script MATLAB files for graphical representation and further numerical analysis. The computer program needs simple data and it is valuable especially when a large number of burnup analyses are required quickly. (authors)

  15. Burnup degree measuring device for spent fuel

    International Nuclear Information System (INIS)

    Doi, Hideo; Imaizumi, Hideki; Endo, Yasumi; Itahara, Kuniyuki.

    1994-01-01

    The present invention provides a small-sized and convenient device for measuring a burnup degree of spent fuels, which can be installed without remodelling an existent fuel storage pool. Namely, a gamma-ray detecting portion incorporates a Cd-Te detector for measuring intensity ratio of gamma-rays. A neutron detecting portion incorporates a fission counter tube. The Cd-Te detector comprises a neutron shielding member for reducing radiation damages and a background controlling plate for reducing low energy gamma-rays entering from a collimator. Since the Cd-Td detector for use in a gamma-ray spectroscopy can be used at a normal temperature and can measure even a relatively strong radiation field, it can measure the intensity of gamma-rays from Cs-137 and Cs-134 in spent fuels accurately at a resolving power of less than 10 keV. Further, in a case where a cooling period is less than one year, gamma-rays from Rh-106 and Nb-95 can also be measured. (I.S.)

  16. Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

  17. Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H.

    2005-01-01

    Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

  18. Water Contact Angle Dependence with Hydroxyl Functional Groups on Silica Surfaces under CO2 Sequestration Conditions.

    Science.gov (United States)

    Chen, Cong; Zhang, Ning; Li, Weizhong; Song, Yongchen

    2015-12-15

    Functional groups on silica surfaces under CO2 sequestration conditions are complex due to reactions among supercritical CO2, brine and silica. Molecular dynamics simulations have been performed to investigate the effects of hydroxyl functional groups on wettability. It has been found that wettability shows a strong dependence on functional groups on silica surfaces: silanol number density, space distribution, and deprotonation/protonation degree. For neutral silica surfaces with crystalline structure (Q(3), Q(3)/Q(4), Q(4)), as silanol number density decreases, contact angle increases from 33.5° to 146.7° at 10.5 MPa and 318 K. When Q(3) surface changes to an amorphous structure, water contact angle increases 20°. Water contact angle decreases about 12° when 9% of silanol groups on Q(3) surface are deprotonated. When the deprotonation degree increases to 50%, water contact angle decreases to 0. The dependence of wettability on silica surface functional groups was used to analyze contact angle measurement ambiguity in literature. The composition of silica surfaces is complicated under CO2 sequestration conditions, the results found in this study may help to better understand wettability of CO2/brine/silica system.

  19. The work of the task group of committee 2 of ICRP on age-dependent dosimetry

    International Nuclear Information System (INIS)

    Stather, J.W.; Kaul, A.; Metivier, H.

    1996-01-01

    With the accident at Chernobyl and developing concern in regard to the consequences of discharging radionuclides into the environment has come increasing awareness of the need to assess radiation doses to all age groups in the population. In 1987, ICRP set up a Task Group of Committee 2 on Age-dependent Dosimetry with the responsibility for calculating internationally agreed dose coefficients for members of the public. This covered the calculation and ingestion, as well as doses to the embryo and fetus from intakes of radionuclides by the mother. This paper reviews the programme of work.(authors). 17 refs., 6 tabs

  20. Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

    Directory of Open Access Journals (Sweden)

    Gajda Paweł

    2016-01-01

    Full Text Available Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom – FIMA. Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/tHM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

  1. High burnup fuel onset conditions in dry storage. Prediction of EOL rod internal pressure

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L.E.

    2015-07-01

    During dry storage, cladding resistance to failure can be affected by several degrading mechanisms like creep or hydrides radial reorientation. The driving force of these effects is the stress at which the cladding is submitted. The maximum stress in the cladding is determined by the end-of-reactor-life (EOL) rod internal pressure, PEOL, at the maximum temperature attained during dry storage. Thus, PEOL sets the initial conditions of storage for potential time-dependent changes in the cladding. Based on FRAPCON-3.5 calculations, the aim of this work is to analyse the PEOL of a PWR fuel rod irradiated to burnups greater than 60 GWd/tU, where limited information is available. In order to be conservative, demanding irradiation histories have been used with a peak linear power of 44 kW/m. FRAPCON-3.5 results show an increasing exponential trend of PEOL with burnup, from which a simple correlation has been derived. The comparison with experimental data found in the literature confirms the enveloping nature of the predicted curve. Based on that, a conservative prediction of cladding stress in dry storage has been obtained. The comparison with a critical stress threshold related to hydrides embrittlement seems to point out that this issue should not be a concern at burnups below 65 GWd/tU. (Author)

  2. Effects of burnup on fission product release and implications for severe fuel damage events

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Cronenberg, A.W.; Carboneau, M.L.

    1984-01-01

    Xe, Kr, and I fission-product release data from (a) Halden tests where release in intact rods was measured during irradiation at burnups to 18,000 MWd/t and fuel temperatures of 800 to 1800 0 K, and (b) Power Burst Facility (PBF) tests where trace-irradiated fuel (approx. = 90 MWd/t) was driven to temperatures of >2400 0 K and fuel liquefaction occurred are discussed and related to fuel morphology. Results from both indicate that the fission-product morphology and fuel restructuring govern release behavior. The Halden tests show low release at beginning of life with a 10-fold increase at burnups in excess of 10,000 MWd/t, due to the development of grain boundary interlinkage at higher burnups. Such dependence of release on morphology characteristics is consistent with findings from the PBF tests, where for trace-irradiated fuel, the absence of interlinkage accounts for the low release rates observed during initial fuel heatup, with subsequent enhanced Xe, Kr, and I release via liquefaction or quench-induced destruction of the grain structure. Morphology is also shown to influence the chemical release form of I and Cs fission products

  3. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

    2014-06-15

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  4. SWAT4.0 - The integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

    International Nuclear Information System (INIS)

    Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki

    2015-03-01

    There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0. (author)

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

    International Nuclear Information System (INIS)

    Conde, J.M.; Recio, M.

    1995-01-01

    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

  6. Anchor-dependent lipofection with non-glycerol based cytofectins containing single 2-hydroxyethyl head groups.

    Science.gov (United States)

    Venkata Srilakshmi, Gollapudi; Sen, Joyeeta; Chaudhuri, Arabinda; Ramadas, Yerramsetti; Madhusudhana Rao, Nalam

    2002-02-15

    Detailed structure-activity investigations aimed at probing the anchor chain length dependency for glycerol-based lipofectins have been reported previously. Herein, we report on the first detailed investigation on the anchor-dependent transfection biology of non-glycerol based simple monocationic cytofectins containing single 2-hydroxyethyl head group functionality using 11 new structural analogs of our previously published first generation of non-glycerol based transfection lipids (lipids 1-11). The C-14 and C-16 analogs of DOMHAC (lipids 4 and 5, respectively) were found to be remarkably efficient in transfecting COS-1 cells. In addition, the present anchor-dependency investigation also revealed that the C-14 analog of DOHEMAB (lipid 10) is significantly efficient in transfecting both COS-1 and NIH3T3 cells. Our results also indicate that too strong lipid-DNA interactions might result in weaker transfection for non-glycerol based cationic lipids. In summary, the anchor-dependence investigations presented here convincingly demonstrate that non-glycerol based cationic lipids containing a single hydroxyethyl head group and hydrophobic C-14 or C-16 anchors are promising non-toxic cationic transfection lipids for future use in liposomal gene delivery.

  7. Nonperturbative renormalization-group approach preserving the momentum dependence of correlation functions

    Science.gov (United States)

    Rose, F.; Dupuis, N.

    2018-05-01

    We present an approximation scheme of the nonperturbative renormalization group that preserves the momentum dependence of correlation functions. This approximation scheme can be seen as a simple improvement of the local potential approximation (LPA) where the derivative terms in the effective action are promoted to arbitrary momentum-dependent functions. As in the LPA, the only field dependence comes from the effective potential, which allows us to solve the renormalization-group equations at a relatively modest numerical cost (as compared, e.g., to the Blaizot-Mendéz-Galain-Wschebor approximation scheme). As an application we consider the two-dimensional quantum O(N ) model at zero temperature. We discuss not only the two-point correlation function but also higher-order correlation functions such as the scalar susceptibility (which allows for an investigation of the "Higgs" amplitude mode) and the conductivity. In particular, we show how, using Padé approximants to perform the analytic continuation i ωn→ω +i 0+ of imaginary frequency correlation functions χ (i ωn) computed numerically from the renormalization-group equations, one can obtain spectral functions in the real-frequency domain.

  8. Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

    International Nuclear Information System (INIS)

    Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok

    2015-01-01

    The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235 U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233 U, 242 Pu, 150 Nd, and 133 Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code

  9. Systemization of burnup sensitivity analysis code (2) (Contract research)

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2008-08-01

    Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion

  10. Numerical solution of the point reactor kinetics equations with fuel burn-up and temperature feedback

    International Nuclear Information System (INIS)

    Tashakor, S.; Jahanfarnia, G.; Hashemi-Tilehnoee, M.

    2010-01-01

    Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during the reactor operation. The variation of reactivity, temperature, and maximum power with time are compared with the predictions by other methods.

  11. Accuracy assessment of a new Monte Carlo based burnup computer code

    International Nuclear Information System (INIS)

    El Bakkari, B.; ElBardouni, T.; Nacir, B.; ElYounoussi, C.; Boulaich, Y.; Meroun, O.; Zoubair, M.; Chakir, E.

    2012-01-01

    Highlights: ► A new burnup code called BUCAL1 was developed. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► Validation of BUCAL1 was done by code to code comparison using VVER-1000 LEU Benchmark Assembly. ► Differences from BM value were found to be ± 600 pcm for k ∞ and ±6% for the isotopic compositions. ► The effect on reactivity due to the burnup of Gd isotopes is well reproduced by BUCAL1. - Abstract: This study aims to test for the suitability and accuracy of a new home-made Monte Carlo burnup code, called BUCAL1, by investigating and predicting the neutronic behavior of a “VVER-1000 LEU Assembly Computational Benchmark”, at lattice level. BUCAL1 uses MCNP tally information directly in the computation; this approach allows performing straightforward and accurate calculation without having to use the calculated group fluxes to perform transmutation analysis in a separate code. ENDF/B-VII evaluated nuclear data library was used in these calculations. Processing of the data library is performed using recent updates of NJOY99 system. Code to code comparisons with the reported Nuclear OECD/NEA results are presented and analyzed.

  12. The criterion-related validity of the Northwick Park Dependency Score as a generic nursing dependency instrument for different rehabilitation patient groups

    NARCIS (Netherlands)

    Plantinga, E.; Tiesinga, L. J.; van der Schans, C. P.; Middel, B.

    2006-01-01

    Objective: To investigate the criterion or concurrent validity of the Northwick Park Dependency Score (NPDS) for determining nursing dependence in different rehabilitation groups, with the Barthel Index (BI) and the Care Dependency Scale (C D S). Design: Cross-sectional study. Setting: Centre for

  13. Burn-up credit in criticality safety of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

    2014-12-15

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

  14. ZZ PWR-AXBUPRO-GKN, Measured Axial Burnup Profiles, NPP Neckarewstheim

    International Nuclear Information System (INIS)

    Neuber, Jens-Christian; Lamprecht, Thomas

    1999-01-01

    Description or function: PWR-AXBUPRO-GKN12 contains Axial Burnup Shapes released by Siemens AG Power Generation Group. It contains data sets relative to following NPP and initial enrichment: - NPP Neckarwestheim 1, Fuel assemblies with an initial enrichment of 3.5 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 3.5 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 3.8 wt.-% 235-U; - NPP Neckarwestheim 2, Fuel assemblies with an initial enrichment of 4.0 wt.-% 235-U. In each of these files the axial shapes are listed one after the other. Each shape is characterised by: - the number of the cycle; - the number of the fuel assembly. - axial burnup shape characteristics: - height (in centimeter) of the nodes with respect to the active length of the fuel assemblies normalized to the cold, unirradiated state. - the nodal burnup (in MWd/kg U). - fuel assembly design data as well as core geometry and operating data pertinent to depletion calculations: NPP Neckarwestheim 1 (GKN1) - (square pitch lattice 15X15 - thermal Power 2497 MW)thermal Power 2497 MW); More than 700 EOC axial shapes from cycle 18 up. From cycle 18 to cycle 20 a change from an Out-In-Loading to an In-Out-Loading has taken place. Fuel assemblies up to number 1093 have spacer grids made of Inconel, whereas all the fuel assemblies from number 1094 up have spacer grids made of Zircaloy. Discharge burnups range from: 9.7 to 52.8 MWd/kg. NPP Neckarwestheim 2 (GKN2) - (square pitch lattice 18X18 - thermal Power 3850 MW) more than 500 EOC axial shapes from cycle 5 up: - More than 170 shapes for an initial fuel enrichment of 3.5 wt.-% 235-U, discharge burnup ranges from 16.3 to 44.4 MWd/kg; - more than 170 shapes for an initial fuel enrichment of 3.8 wt.-% 235-U, discharge burnup ranges from 14.0 to 52.8 MWd/kg; - more than 180 shapes for an initial fuel enrichment of 4.0 wt.-% 235-U. discharge burnup ranges from 15.5 to 48.9 MWd/kg. PWR AXBUPRO

  15. Context-dependent effects of background colour in free recall with spatially grouped words.

    Science.gov (United States)

    Sakai, Tetsuya; Isarida, Toshiko K; Isarida, Takeo

    2010-10-01

    Three experiments investigated context-dependent effects of background colour in free recall with groups of items. Undergraduates (N=113) intentionally studied 24 words presented in blocks of 6 on a computer screen with two different background colours. The two background colours were changed screen-by-screen randomly (random condition) or alternately (alternation condition) during the study period. A 30-second filled retention interval was imposed before an oral free-recall test. A signal for free recall was presented throughout the test on one of the colour background screens presented at study. Recalled words were classified as same- or different-context words according to whether the background colours at study and test were the same or different. The random condition produced significant context-dependent effects, whereas the alternation condition showed no context-dependent effects, regardless of whether the words were presented once or twice. Furthermore, the words presented on the same screen were clustered in recall, whereas the words presented against the same background colour but on different screens were not clustered. The present results imply: (1) background colours can cue spatially massed words; (2) background colours act as temporally local context; and (3) predictability of the next background colour modulates the context-dependent effect.

  16. Achieving High Burnup Targets With Mox Fuels: Techno Economic Implications

    International Nuclear Information System (INIS)

    Clement Ravi Chandar, S.; Sivayya, D.N.; Puthiyavinayagam, P.; Chellapandi, P.

    2013-01-01

    For a typical MOX fuelled SFR of power reactor size, Implications due to higher burnup have been quantified. Advantages: – Improvement in the economy is seen upto 200 GWd/ t; Disadvantages: – Design changes > 150 GWd/ t bu; – Need for 8/ 16 more fuel SA at 150/ 200 GWd/ t bu; – Higher enrichment of B 4 C in CSR/ DSR at higher bu; – Reduction in LHR may be required at higher bu; – Structural material changes beyond 150 GWd/ t bu; – Reprocessing point of view-Sp Activity & Decay heat increase. Need for R & D is a must before increasing burnup. bu- refers burnup. Efforts to increase MOX fuel burnup beyond 200 GWd/ t may not be highly lucrative; • MOX fuelled FBR would be restricted to two or four further reactors; • Imported MOX fuelled FBRs may be considered; • India looks towards launching metal fuel FBRs in the future. – Due to high Breeding Ratio; – High burnup capability

  17. Status of burnup credit implementation and research in Switzerland

    International Nuclear Information System (INIS)

    Grimm, P.

    2001-01-01

    Burnup credit has recently been approved by the Swiss licensing authority for the spent-fuel storage pool of a PWR plant for fuel exceeding the originally licensed initial enrichment. The criticality safety assessment is based on a configuration consisting of a small number (approximately a reload batch) of fresh assemblies surrounded by assemblies having a burnup corresponding to the minimum value in the top 1 m section after one cycle of irradiation. The allowable initial enrichment in this configuration is about 0.5% higher than for all fresh fuel. A central storage facility for all types of radioactive wastes from Switzerland, including cask storage of spent fuel assemblies is being commissioned presently. The first applications for licenses for casks to be used in this facility have been submitted. Credit for burnup has not been requested in these applications (conforming to the original licenses of the casks in their countries of origin), but utilities are interested in burnup credit for fuel with higher initial enrichments. Reactivity worth measurements as well as chemical assays of spent fuel samples in the LWR-PROTEUS facility at PSI are in detailed planning currently. The experiments, scheduled to start in 2001, will be performed in cooperation with the Swiss utilities and their fuel vendors. Although the focus of interest of these partners is on validation of in-core fuel management tools, the same experiments are also applicable to burnup credit, and contacts with further potential partners interested in this field are underway. (author)

  18. Advances in Metallic Fuels for High Burnup and Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

    2016-10-01

    Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

  19. Mining approximate temporal functional dependencies with pure temporal grouping in clinical databases.

    Science.gov (United States)

    Combi, Carlo; Mantovani, Matteo; Sabaini, Alberto; Sala, Pietro; Amaddeo, Francesco; Moretti, Ugo; Pozzi, Giuseppe

    2015-07-01

    Functional dependencies (FDs) typically represent associations over facts stored by a database, such as "patients with the same symptom get the same therapy." In more recent years, some extensions have been introduced to represent both temporal constraints (temporal functional dependencies - TFDs), as "for any given month, patients with the same symptom must have the same therapy, but their therapy may change from one month to the next one," and approximate properties (approximate functional dependencies - AFDs), as "patients with the same symptomgenerallyhave the same therapy." An AFD holds most of the facts stored by the database, enabling some data to deviate from the defined property: the percentage of data which violate the given property is user-defined. According to this scenario, in this paper we introduce approximate temporal functional dependencies (ATFDs) and use them to mine clinical data. Specifically, we considered the need for deriving new knowledge from psychiatric and pharmacovigilance data. ATFDs may be defined and measured either on temporal granules (e.g.grouping data by day, week, month, year) or on sliding windows (e.g.a fixed-length time interval which moves over the time axis): in this regard, we propose and discuss some specific and efficient data mining techniques for ATFDs. We also developed two running prototypes and showed the feasibility of our proposal by mining two real-world clinical data sets. The clinical interest of the dependencies derived considering the psychiatry and pharmacovigilance domains confirms the soundness and the usefulness of the proposed techniques. Copyright © 2014 Elsevier Ltd. All rights reserved.

  20. Reference-Dependent Aggregation in Multi-AttributeGroup Decision-Making

    Directory of Open Access Journals (Sweden)

    Jianwei Gao

    2017-03-01

    Full Text Available To characterize the influence of decision makers’ psychological factors on the group decisionprocess, this paper develops a new class of aggregation operators based on reference-dependentutility functions (RUs in multi-attribute group decision analysis. We consider two types of RUs:S-shaped, representing decision makers who are risk-seeking for relative losses, and non-S-shaped,representing those that are risk-averse for relative losses. Based on these RUs, we establish twonew classes of reference-dependent aggregation operators; we study their properties and showthat their generality covers a number of existing aggregation operators. To determine the optimalweights for these aggregation operators, we construct an attribute deviation weight model and adecision maker (DM deviation weight model. Furthermore, we develop a new multi-attribute groupdecision-making (MAGDM approach based on these RU aggregation operators and weight models.Finally, numerical examples are given to illustrate the application of the approach.

  1. Possible Involvement of Hydrosulfide in B12-Dependent Methyl Group Transfer

    Directory of Open Access Journals (Sweden)

    John I. Toohey

    2017-04-01

    Full Text Available Evidence from several fields of investigation lead to the hypothesis that the sulfur atom is involved in vitamin B12-dependent methyl group transfer. To compile the evidence, it is necessary to briefly review the following fields: methylation, the new field of sulfane sulfur/hydrogen sulfide (S°/H2S, hydrosulfide derivatives of cobalamins, autoxidation of hydrosulfide radical, radical S-adenosylmethionine methyl transfer (RSMT, and methionine synthase (MS. Then, new reaction mechanisms for B12-dependent methyl group transfer are proposed; the mechanisms are facile and overcome difficulties that existed in previously-accepted mechanisms. Finally, the theory is applied to the effect of S°/H2S in nerve tissue involving the “hypomethylation theory” that was proposed 50 years ago to explain the neuropathology resulting from deficiency of vitamin B12 or folic acid. The conclusions are consistent with emerging evidence that sulfane sulfur/hydrogen sulfide may be beneficial in treating Alzheimer’s disease.

  2. Modelling of some high burnup phenomena in nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, K; Lindstroem, F; Massih, A R [ABB Atom AB, Vaesteraas (Sweden)

    1997-08-01

    In this paper the results of some modelling efforts carried out by ABB Atom to describe certain light water reactor fuel high burnup effects are presented. In particular the degradation of fuel thermal conductivity with burnup and its impact on fuel temperature is briefly discussed. The formation of a porous rim and its effect on a thermal fission gas release has been modelled and the model has been used to predict the release of pressurized water reactor fuel rods that were operated at low power densities. Furthermore, a mathematical model which combines the diffusion and re-solution controlled thermal release with grain boundary movement has been briefly described. The model is used to compare release with diffusion only and release caused by diffusion and grain boundary sweeping (due to grain growth). Finally, analytical expressions are obtained for the calculation of fuel stoichiometry as a function of burnup. (author). 20 refs, 10 figs, 1 tab.

  3. Validation issues for depletion and criticality analysis in burnup credit

    International Nuclear Information System (INIS)

    Parks, C.V.; Broadhead, B.L.; Dehart, M.D.; Gauld, I.C.

    2001-01-01

    This paper reviews validation issues associated with implementation of burnup credit in transport, dry storage, and disposal. The issues discussed are ones that have been identified by one or more constituents of the United States technical community (national laboratories, licensees, and regulators) that have been exploring the use of burnup credit. There is not necessarily agreement on the importance of the various issues, which sometimes is what creates the issue. The broad issues relate to the paucity of available experimental data (radiochemical assays and critical experiments) covering the full range and characteristics of spent nuclear fuel in away-from-reactor systems. The paper will also introduce recent efforts initiated at Oak Ridge National Laboratory (ORNL) to provide technical information that can help better assess the value of different experiments. The focus of the paper is on experience with validation issues related to use of burnup credit for transport and dry storage applications. (author)

  4. Preparation of higher-actinide burnup and cross section samples

    International Nuclear Information System (INIS)

    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 241 Am and 244 Cm in the forms of Am 2 O 3 , Cm 2 O 3 , and Am 6 Cm(RE) 7 O 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

  5. Fission gas release from fuels at high burnup

    International Nuclear Information System (INIS)

    Kauffmann, Yves; Pointud, M.L.; Vignesoult, Nicole; Atabek, Rosemarie; Baron, Daniel.

    1982-04-01

    Determinations of residual gas concentrations by heating and by X microanalysis were respectively carried out on particles (TANGO program) and on sections of fuel rods, perfectly characterized as to fabrication and irradiation history. A threshold release temperature of 1250 0 C+-100 0 C was determined irrespective of the type of oxide and the irradiation history in the 18,000-45,000 MWdt -1 (U) specific burnup field. The overall analyses of gas released from the fuel rods show that, in the PWR operating conditions, the fraction released remains less than 1% up to a mean specific burnup of 35000 MWdt -1 (U). The release of gases should not be a limiting factor in the increase of specific burnups [fr

  6. Increased fuel burn-up and fuel cycle equilibrium

    International Nuclear Information System (INIS)

    Debes, M.

    2001-01-01

    Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

  7. Role of measurement systems in burnup credit operations

    International Nuclear Information System (INIS)

    Ewing, R.I.; Sanders, T.L.

    1991-01-01

    Spent fuel transport casks designed using burnup credit have increased payloads that may greatly reduce the number of shipments required to transport spent fuel from reactor sites to repositories. Burnup credit is obtained by applying the reduced reactivity of spent fuel to considerations of nuclear criticality in the design of transport casks. Although it does not appear to be possible to directly measure the criticality of spent fuel assemblies, measurements can be employed to ensure that the only assemblies loaded into a cask have the characteristics appropriate to that cask design. An effective on-site measurement system must be matched to the characteristics of the spent fuel cask design and to the inventory of spent fuel. For operation reasons the system should be simple, accurate, efficient, and easily calibrated. This paper is part of a study to examine the effects of the spent fuel inventory in the U.S. on the selection of measurement systems useful in burnup credit operations

  8. Burnup credit implementation plan and preparation work at JAERI

    International Nuclear Information System (INIS)

    Nomura, Y.; Itahara, K.

    2001-01-01

    Application of the burnup credit concept is considered to be very effective to the design of spent fuel transport and storage facilities. This technology is all the more important when considering construction of the intermediate spent fuel storage facility, which is to be commissioned by 2010 due to increasing amount of accumulated spent fuel in Japan. Until reprocessing and recycling all the spent fuel arising, they will be stored as an energy stockpile until such time as they can be reprocessed. On the other hand, the burnup credit has been partly taken into account for the spent fuel management at Rokkasho Reprocessing Plant, which is to be commissioned in 2005. They have just finished the calibration tests for their burnup monitor with initially accepted several spent fuel assemblies. Because this monitoring system is employed with highly conservative safety margin, it is considered necessary to develop the more rational and simplified method to confirm burnup of spent fuel. A research program has been instituted to improve the present method employed at the spent fuel management system for the Spent Fuel Receiving and Storage Pool of Rokkasho Reprocessing Plant. This program is jointly performed by Japan Nuclear Fuel Limited (JNFL) and JAERI.This presentation describes the current status of spent fuel accumulation discharged from PWR and BWR in Japan and the recent incentive to introduce burnup credit into design of spent fuel storage and transport facilities. This also includes the content of the joint research program initiated by JNFL and JAERI. The relevant study has been continued at JAERI. The results by these research programs will be included in the Burnup Credit Guide Original Version compiled by JAERI. (author)

  9. MIRU-VNTR allelic variability depends on Mycobacterium bovis clonal group identity.

    Science.gov (United States)

    Hauer, Amandine; Michelet, Lorraine; De Cruz, Krystel; Cochard, Thierry; Branger, Maxime; Karoui, Claudine; Henault, Sylvie; Biet, Franck; Boschiroli, María Laura

    2016-11-01

    The description of the population of M. bovis strains circulating in France from 1978 to 2013 has highlighted the discriminating power of the MLVA among predominant spoligotype groups. In the present study we aimed to characterize clonal groups via MLVA and to better understand the strain's population structure. MLVA was performed with eight MIRU-VNTR loci, most of them defined by the Venomyc European consortium. The discriminatory index of each MLVA loci was calculated for SB0120, SB0134, SB0121 and the "F4-family", the main spoligotype groups in France. Differences in global DI per spoligotype, but also by locus within each spoligotype, were observed, which strongly suggest the clonal complex nature of these major groups. These MLVA results were compared to those of other European countries where strain collections had been characterized (Spain, Portugal, Italy, Northern Ireland and Belgium). Overall, QUB 3232 and ETR D are respectively the most and the least discriminative loci, regardless of the strains geographical origin. However, marked DI differences are observed in the rest of the MIRU-VNTR loci, again highlighting that strain genetic variability in a country depends on the dominant existing clonal complexes. A web application for M. bovis, including spoligotyping and MIRU-VNTR typing data, was developed to allow inter-laboratory comparison of field isolates. In conclusion, combination of typing methods is required for M. bovis optimum discrimination and differentiation of groups of strains. Thus, the loci employed for MLVA in a country should be those which are the most discriminative for the clonal complexes which characterize their M. bovis population. Copyright © 2016 Elsevier B.V. All rights reserved.

  10. Consequences of the increase of burnup on the fuel

    International Nuclear Information System (INIS)

    Melin, P.; Lavoine, O.; Houdaille, B.

    1986-04-01

    The examinations carried out on the FRAGEMA fuel of EDF reactors show its good behavior in service. The results of research and development programs developed by EDF, FGA and the CEA show that this fuel can be irradiated up to a high burnup, and allow to point out the axies of research to improve still the performance of the product in a more and more soliciting environment (increase of power and burnup coupled with load following). Among the solutions considered, there are the design and fabrication adjustments (geometry, initial pressurization), more fundamental changes concerning fuel cans and fuel pellets, which need still research and development programs [fr

  11. Fission gas release from fuel at high burnup

    International Nuclear Information System (INIS)

    Meyer, R.O.; Beyer, C.E.; Voglewede, J.C.

    1978-03-01

    The release of fission gas from fuel pellets at high burnup is reviewed in the context of the safety analysis performed for reactor license applications. Licensing actions are described that were taken to correct deficient gas release models used in these safety analyses. A correction function, which was developed by the Nuclear Regulatory Commission staff and its consultants, is presented. Related information, which includes some previously unpublished data, is also summarized. The report thus provides guidance for the analysis of high burnup gas release in licensing situations

  12. Burnup measurements with the Los Alamos fork detector

    International Nuclear Information System (INIS)

    Bosler, G.E.; Rinard, P.M.

    1991-01-01

    The fork detector system can determine the burnup of spent-fuel assemblies. It is a transportable instrument that can be mounted permanently in a spent-fuel pond near a loading area for shipping casks, or be attached to the storage pond bridge for measurements on partially raised spent-fuel assemblies. The accuracy of the predicted burnup has been demonstrated to be as good as 2% from measurements on assemblies in the United States and other countries. Instruments have also been developed at other facilities throughout the world using the same or different techniques, but with similar accuracies. 14 refs., 2 figs., 2 tabs

  13. Isotopic biases for actinide-only burnup credit

    International Nuclear Information System (INIS)

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

    1997-01-01

    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

  14. Tag gas burnup based on three-dimensional FTR analysis

    International Nuclear Information System (INIS)

    Kidman, R.B.

    1976-01-01

    Flux spectra from a three-dimensional diffusion theory analysis of the Fast Test Reactor (FTR) are used to predict gas tag ratio changes, as a function of exposure, for each FTR fuel and absorber subassembly plenum. These flux spectra are also used to predict Xe-125 equilibrium activities in absorber plena in order to assess the feasibility of using Xe-125 gamma rays to detect and distinguish control rod failures from fuel rod failures. Worst case tag burnup changes are used in conjunction with burnup and mass spectrometer uncertainties to establish the minimum spacing of tags which allows the tags to be unambiguously identified

  15. Nuclear fuel burn-up economy; Ekonomija izgaranja nuklearnog goriva

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1984-07-01

    In the period 1981-1985, for the needs of Utility Organization, Beograd, and with the support of the Scientific Council of SR Srbija, work has been performed on the study entitled 'Nuclear Fuel Burn-up Economy'. The forst [phase, completed during the year 1983 comprised: comparative analysis of commercial NPP from the standpoint of nuclear fuel requirements; development of methods for fuel burn-up analysis; specification of elements concerning the nuclear fuel for the tender documentation. The present paper gives the short description of the purpose, content and results achieved in the up-to-now work on the study. (author)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

    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

  17. A regime showing anomalous triton burnup in JET

    International Nuclear Information System (INIS)

    Conroy, S.; Jarvis, O.N.; Sadler, G.; Pillon, M.

    1990-01-01

    Measurements of triton burnup made at JET in 1989 are in good agreement with a simple classical model of the triton slowing down, for the majority of discharges. For discharges with a long slowing down time (greater than 2 seconds), a much reduced burnup has been observed, suggesting that the tritons undergo diffusion with a diffusion constant of 0.10 m 2 s -1 . Also, the experimental 14 MeV neutron yield is 30% lower than expected for Beryllium limiter discharges. (author) 4 refs., 3 figs

  18. 75 FR 27121 - Interim Final Rules for Group Health Plans and Health Insurance Issuers Relating to Dependent...

    Science.gov (United States)

    2010-05-13

    ... 45 CFR Parts 144, 146, and 147 Group Health Plans and Health Insurance Issuers Relating to Dependent... 144, 146, and 147 RIN 0991-AB66 Interim Final Rules for Group Health Plans and Health Insurance... requirements for group health plans and health insurance issuers in the group and individual markets under...

  19. The oxidative costs of reproduction are group-size dependent in a wild cooperative breeder.

    Science.gov (United States)

    Cram, Dominic L; Blount, Jonathan D; Young, Andrew J

    2015-11-22

    Life-history theory assumes that reproduction entails a cost, and research on cooperatively breeding societies suggests that the cooperative sharing of workloads can reduce this cost. However, the physiological mechanisms that underpin both the costs of reproduction and the benefits of cooperation remain poorly understood. It has been hypothesized that reproductive costs may arise in part from oxidative stress, as reproductive investment may elevate exposure to reactive oxygen species, compromising survival and future reproduction and accelerating senescence. However, experimental evidence of oxidative costs of reproduction in the wild remains scarce. Here, we use a clutch-removal experiment to investigate the oxidative costs of reproduction in a wild cooperatively breeding bird, the white-browed sparrow weaver, Plocepasser mahali. Our results reveal costs of reproduction that are dependent on group size: relative to individuals in groups whose eggs were experimentally removed, individuals in groups that raised offspring experienced an associated cost (elevated oxidative damage and reduced body mass), but only if they were in small groups containing fewer or no helpers. Furthermore, during nestling provisioning, individuals that provisioned at higher rates showed greater within-individual declines in body mass and antioxidant protection. Our results provide rare experimental evidence that reproduction can negatively impact both oxidative status and body mass in the wild, and suggest that these costs can be mitigated in cooperative societies by the presence of additional helpers. These findings have implications for our understanding of the energetic and oxidative costs of reproduction, and the benefits of cooperation in animal societies. © 2015 The Authors.

  20. Dependence of transmittance and group index on the coupling strength between constituents of a metamaterial

    International Nuclear Information System (INIS)

    Thuy Vu, Tran Thanh; Lee, Young Pak; Nguyen, Thanh Tung; Rhee, Joo Yull; Vu, Dinh Lam

    2011-01-01

    Recent studies on the coupling effects between constituent elements of metamaterials have opened up a new gateway to many fascinating electromagnetic properties and functionalities that cannot be explained by the uncoupled point of view. In this work, we numerically investigated, in a THz regime, the coupling between a cut wire and a split-ring resonator, which gives rise to an interesting phenomenon—the so-called electromagnetically induced transparency-like effect. The trade-off between the maximum transmittance of the transmission window and the group index, which depends on the coupling strength between constituent elements, was systematically studied. Furthermore, by characterizing this trade-off by the transmittance-delay product (figure of merit), a criterion for slow-light applications was provided

  1. REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR

    International Nuclear Information System (INIS)

    Boettcher, W.; Schmidt, E.

    1969-01-01

    1 - Nature of physical problem solved: REFLOS is a programme for the evaluation of fuel-loading schemes in heavy water moderated reactors. The problems involved in this study are: a) Burn-up calculation for the reactor cell. b) Determination of reactivity behaviour, power distribution, attainable burn-up for both the running-in period and the equilibrium of a 3-dimensional heterogeneous reactor model; investigation of radial fuel movement schemes. c) Evaluation of mass flows of heavy atoms through the reactor and fuel cycle costs for the running-in, the equilibrium, and the shut down of a power reactor. If the subroutine for treating the reactor cell were replaced by a suitable routine, other reactors with weakly absorbing moderators could be analyzed. 2 - Method of solution: Nuclear constants and isotopic compositions of the different fuels in the reactor are calculated by the cell-burn-up programme and tabulated as functions of the burn-up rate (MWD/T). Starting from a known state of the reactor, the 3-dimensional heterogeneous reactor programme (applying an extension of the technique of Feinberg and Galanin) calculates reactivity and neutron flux distribution using one thermal and one or two fast neutron groups. After a given irradiation time, the new state of the reactor is determined, and new nuclear constants are assigned to the various defined locations in the reactor. Reloading of fuel may occur if the prescribed life of the reactor is reached or if the effective multiplication factor or the power form factor falls below a specified level. The scheme of reloading to be carried out is specified by a load vector, giving the number of channels to be discharged, the kind of movement from one to another channel and the type of fresh fuel to be charged for each single reloading event. After having determined the core states characterizing the equilibrium period, and having decided the fuel reloading scheme for the running-in period of the reactor life, the fuel

  2. Effects of a Dependent Group Contingency on the Verbal Interactions of Middle School Students with Emotional Disturbance

    OpenAIRE

    Hansen, S.; Lignugaris/Kraft, Benjamin

    2005-01-01

    This study examined the effects of a dependent group contingency to increase positive verbal interactions among nine middle school-aged males in a self-contained classroom. Prior to implementing the group contingency, the students received social skills instruction. A withdrawal experimental design was then used to evaluate the effects of the dependent group contingency. Students' positive verbal interactions increased and negative verbal interactions decreased with the implementation of the ...

  3. COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

    International Nuclear Information System (INIS)

    Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.

    1964-01-01

    1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100

  4. Burnup measurements at the RECH-1 research reactor

    International Nuclear Information System (INIS)

    Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

    2002-01-01

    The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

  5. The Gd-isotopic fuel for high burnup in PWR's

    International Nuclear Information System (INIS)

    Dias, Marcio Soares; Mattos, João Roberto L. de; Andrade, Edison Pereira de

    2017-01-01

    Today, the discussion about the high burnup fuel is beyond the current fuel enrichment licensing and burnup limits. Licensing issues and material/design developments are again key features in further development of the LWR fuel design. Nevertheless, technological and economical solutions are already available or will be available in a short time. In order to prevent the growth of the technological gap, Brazil's nuclear sector needs to invest in the training of new human resources, in the access to international databases, and in the upgrading existing infrastructure. Experimental database and R&D infrastructure are essential components to support the autonomous development of Brazilian Nuclear Reactors, promoting the development of national technologies. The (U,Gd)O_2 isotopic fuel proposed by the CDTN's staff solve two main issues in the high burnup fuel, which are (1) the peak of reactivity resulting from the Gd-157 fast burnup, and (2) the peak of temperature in the (U,Gd)O_2 nuclear fuel resulting from detrimental effects in the thermal properties for gadolinia additions higher than 2%. A sustainable future can be envisaged for the nuclear energy. (author)

  6. Ultrasonic measurement of high burn-up fuel elastic properties

    International Nuclear Information System (INIS)

    Laux, D.; Despaux, G.; Augereau, F.; Attal, J.; Gatt, J.; Basini, V.

    2006-01-01

    The ultrasonic method developed for the evaluation of high burn-up fuel elastic properties is presented hereafter. The objective of the method is to provide data for fuel thermo-mechanical calculation codes in order to improve industrial nuclear fuel and materials or to design new reactor components. The need for data is especially crucial for high burn-up fuel modelling for which the fuel mechanical properties are essential and for which a wide range of experiments in MTR reactors and high burn-up commercial reactor fuel examinations have been included in programmes worldwide. To contribute to the acquisition of this knowledge the LAIN activity is developing in two directions. First one is development of an ultrasonic focused technique adapted to active materials study. This technique was used few years ago in the EdF laboratory in Chinon to assess the ageing of materials under irradiation. It is now used in a hot cell at ITU Karlsruhe to determine the elastic moduli of high burnup fuels from 0 to 110 GWd/tU. Some of this work is presented here. The second on going programme is related to the qualification of acoustic sensors in nuclear environments, which is of a great interest for all the methods, which work, in a hostile nuclear environment

  7. Burnup measurement study and prototype development in HTR-PM

    International Nuclear Information System (INIS)

    Yan Weihua; Zhang Zhao; Xiao Zhigang; Zhang Liguo

    2014-01-01

    In a pebble-bed core which employs the multi-pass scheme, it is mandatory to determine the burnup of each pebble after the pebble has been extracted from the core in order to determine whether its design burnup has been reached or whether it has to be reinserted into the core again. The burnup of the fuel pebbles can be determined by measuring the activity of 137 Cs with an HPGe detector because of their good correspondence, which is independent of the irradiation history in the core. Based on experiments and Geant4 simulation, the correction factor between the fuel and calibration source was derived by using the efficiency transfer method. By optimizing spectrum analysis algorithm and parameters, the relative standard deviation of the 137 Cs activity can be still controlled below 3.0% despite of the presence of interfering peaks. On the foundation of the simulation and experiment research, a complete solution for burnup measurement system in HTR-PM is provided. (authors)

  8. Extended burnup with SEU fuel in Atucha-1 NPP

    International Nuclear Information System (INIS)

    Alvarez, L.; Casario, J.; Fink, J.; Perez, R.; Higa, M.

    2002-01-01

    Atucha-1 is a Pressurized Heavy Water Reactor originally fuelled with natural uranium. Fuel Assemblies consist of 36 fuel rods and the active length is 5300 mm. The total length of the fuel assembly is about 6 m. The average discharge burnup of natural UO 2 fuel is 5900 MWd/tU. After the deregulation of the Argentine electricity market there was an important incentive to reduce the impact of fuel cost on the cost of generation. To keep the competitiveness of the nuclear energy against another sources of electricity it was necessary to reduce the cost of the nuclear fuel. With this objective a program to introduce SEU (0.85 % 235 U) fuel in Atucha-1 was launched in 1993. As a result of this program the average SEU fuel discharge burnup increased to more than 11000 MWd/tU. The first SEU fuels were introduced in Atucha-1 in 1995 and, in the present stage of the program, 71% of core positions are loaded with this type of fuel. This paper describes key aspects of Atucha-1 fuel design and their relevance limiting the burnup extension and shows relevant data regarding the SEU in-reactor performance. At the present time 125 SEU Fuel Assemblies have been irradiated without failures associated with the extended burnup or unfavorable influences on the operation of the power station. (author)

  9. M5TM alloy high burnup behavior and worldwide licensing

    International Nuclear Information System (INIS)

    Mardon, J.P.; Hoffmann, P.B.; Garner, G.L.

    2005-01-01

    The in-reactor behavior of advanced PWR Zirconium alloys at burnups equal to or below licensing limits has been widely reported. Specifically, the advanced alloy M5 has demonstrated impressive improvements over Zircaloy-4 for fuel rod cladding and fuel assembly structural components. To demonstrate superiority of the alloy at burnups beyond current licensing limits, M5 has been operated in PWR at burnups exceeding 71 GWd/tU in the United States and 78 GWd/tU in Europe. Two extensive irradiation programs have been performed in the United States to demonstrate alloy M5 performance beyond current licensing limits. Four M5 TM fuel rods were exposed to four 24-month cycles in a 15x15 reactor beginning in 1995. Additionally, one 17x17 lead assembly containing M5 fuel rods and guide tubes was operated for four 18-month cycles beginning from 1997. Post-irradiation examinations (PIE) performed after all four cycles in the 15x15 demonstration program revealed excellent performance in the licensed burnup and in the high burnup stages of the experience. Examination of the 4th cycle 17x17 assembly will be accomplished in two stages the first of which is scheduled for June 2005. Moreover, several irradiation campaigns have been performed in Europe in order to confirm the excellent M5 in-pile behavior in demanding PWRs irradiation conditions with regard to void fraction, heat flux, lithium content and temperature. Results from the high burnup fuel examinations verify that the excellent performance achieved up to 62 GWd/tU was continued into higher burnup. The results of high burnup PIE campaigns for European and American PWR's are presented in this paper. Measured performance indicators include fuel assembly dimensional stability parameters (assembly length, fuel rod length, assembly bow, fuel rod bow, fuel rod radial creep and spacer grid width), oxidation measurements (fuel rod and guide tube) and hydrogen pick-up data (fuel rod). In the framework of PCI studies, power ramp

  10. Group boundary permeability moderates the effect of a dependency meta-stereotype on help-seeking behaviour.

    Science.gov (United States)

    Zhang, Lange; Kou, Yu; Zhao, Yunlong; Fu, Xinyuan

    2016-08-01

    Previous studies have found that when low-status group members are aware that their in-group is stereotyped as dependent by a specific out-group (i.e. a dependency meta-stereotype is salient), they are reluctant to seek help from the high-status out-group to avoid confirming the negative meta-stereotype. However, it is unclear whether low-status group members would seek more help in the context of a salient dependency meta-stereotype when there is low (vs. high) group boundary permeability. Therefore, we conducted two experiments to examine the moderating effect of permeability on meta-stereotype confirmation with a real group. In study 1, we manipulated the salience of the dependency meta-stereotype, measured participants' perceived permeability and examined their help-seeking behaviour in a real-world task. Participants who perceived low permeability sought more help when the meta-stereotype was salient (vs. not salient), whereas participants who perceived high permeability sought the same amount of help across conditions. In study 2, we manipulated the permeability levels and measured the dependency meta-stereotype. Participants who endorsed a high-dependency meta-stereotype sought more help than participants who endorsed a low-dependency meta-stereotype; this effect was particularly strong in the low-permeability condition. The implications of these results for social mobility and intergroup helping are discussed. © 2015 International Union of Psychological Science.

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

    International Nuclear Information System (INIS)

    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 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 ∼ 65 MWd/kgM. (2) Oxidation of spent fuel up to the U 4 O 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 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 129 I gap inventory for US LWR fuels

  12. Reactivity effect of spent fuel due to spatial distributions for coolant temperature and burnup

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, T.; Yamane, Y. [Nagoya Univ., Dept. of Nuclear Engineering, Nagoya, Aichi (Japan); Suyama, K. [OECD/NEA, Paris (France); Mochizuki, H. [Japan Research Institute, Ltd., Tokyo (Japan)

    2002-03-01

    We investigated the reactivity effect of spent fuel caused by the spatial distributions of coolant temperature and burnup by using the integrated burnup calculation code system SWAT. The reactivity effect which arises from taking account of the spatial coolant temperature distribution increases as the average burnup increases, and reaches the maximum value of 0.69%{delta}k/k at 50 GWd/tU when the burnup distribution is concurrently considered. When the burnup distribution is ignored, the reactivity effect decreases by approximately one-third. (author)

  13. MODRIB - a zero dimensional code for criticality and burn-up of LWR's

    International Nuclear Information System (INIS)

    Gaafar, M.A.; El-Cherif, A.I.

    1980-01-01

    The computer program MODRIB is a zero-dimensional code for calculating criticality and burn-up of light water reactors (LWR's). It is a version of an Italian code RIBOT-2 with an updated cross-section data library. The nuclear constants of MODRIB-code are calculated with a two group scheme (fast and thermal), where the fast group is an average of three fast groups. The code requires as input data essential extensive reactor parameters such as fuel rod radius, clad thickness, fuel enrichment, lattice pitch, water density and temperature etc. A summary of the physical model description and the input-output procedures are given in this report. Selected results of two sample problems are also given for the purpose of checking the validity and reliability of the code. The first is BWR and the second is PWR. The calculation time for a criticality problem with burn-up is about 8 seconds for the first time step and about 3 seconds for each subsequent time step on the ICL-1906 computer facility. The requirements on the memory size is less than 32 K-word. (author)

  14. Burnup calculation with estimated neutron spectrum of JMTR irradiation field. Development of the burnup calculation method for fuel pre-irradiated in the JMTR

    International Nuclear Information System (INIS)

    Okonogi, Kazunari; Nakamura, Takehiko; Yoshinaga, Makio; Hosoyamada, Ryuji

    1999-03-01

    As a series of the pulse irradiation tests with the irradiated fuel, the high-enriched fuel rods pre-irradiated in the JMTR as well as the fuels irradiated in commercial reactors have been irradiated in the NSRR. In the pre-irradiation at the JMTR, the test fuels were placed at the irradiation holes in the reflector region far from the driver core to keep the linear heat generation rate of the test fuel low. Accordingly, neutron energy spectra of the irradiation holes for the test fuels are softened due to the higher moderator ratio than in those of the ordinary LWR core, which causes quite different burnup characteristics. JMTR post irradiation condition corresponds to the pre-test condition in the NSRR. Therefore, proper understanding of the condition is quite important for the precise evaluating the energy deposition and FP generation in the test. Then, neutron spectra at the JMTR irradiation field were evaluated and its effects on the burnup calculation were quantified. Basing on the configuration of the JMTR core in the operation cycle No.85, neutron diffusion calculations of 107 groups were executed in 2-D slab (X-Y) geometry of CITATION of SRAC95 code system, and neutron energy spectra of the irradiation hole for the test fuels were evaluated. Burnup calculations of Test JMN-1 fuel with the estimated neutron energy spectra were performed and the results were compared to both the measurements and calculation results with the PWR and BWR libraries in ORIGEN2 code. SWAT code was used to collapse the 107 groups spectra into 1 group libraries for the ORIGEN2 use. The calculation results for both the generation and depletion of U, Pu and Nd with the JMTR libraries obtained in the present study were in the reasonably good agreement with the measurements, while in the case of calculation with the PWR and BWR libraries in ORIGEN2, the generation of fission products having mass numbers from 105 to 130 and some actinides were overestimated by about 1.5 to 3.5 times

  15. Teachers' Promotion or Inhibition of Children's Aggression Depends on Peer-Group Characteristics.

    Science.gov (United States)

    Peets, Kätlin; Kikas, Eve

    2017-01-01

    Researchers have increasingly started to pay attention to how contextual factors, such as the classroom peer context and the quality of student-teacher interactions, influence children's aggressive behavior. This longitudinal study was designed to examine the degree to which benefits and costs of different teaching practices (child-centered and child-dominated) would be dependent on the initial peer-group composition (aggregate levels of aggression and victimization at the beginning of first grade). Teachers provided ratings of aggression and victimization (N = 523 first-grade students; M age at the beginning of first grade = 7.49 years, SD = 0.52). Information about different teaching practices was obtained via observations. Our results show that whereas child-centered practices are beneficial in high-victimization classrooms, child-dominated practices inhibit the development of aggression in low-victimization classroom contexts. Our findings highlight the importance of moving beyond main-effect models to studying how different contextual influences interact to promote, or inhibit, the development of aggression.

  16. Polycomb Group Protein PHF1 Regulates p53-dependent Cell Growth Arrest and Apoptosis*

    Science.gov (United States)

    Yang, Yang; Wang, Chenji; Zhang, Pingzhao; Gao, Kun; Wang, Dejie; Yu, Hongxiu; Zhang, Ting; Jiang, Sirui; Hexige, Saiyin; Hong, Zehui; Yasui, Akira; Liu, Jun O.; Huang, Haojie; Yu, Long

    2013-01-01

    Polycomb group protein PHF1 is well known as a component of a novel EED-EZH2·Polycomb repressive complex 2 complex and plays important roles in H3K27 methylation and Hox gene silencing. PHF1 is also involved in the response to DNA double-strand breaks in human cells, promotes nonhomologous end-joining processes through interaction with Ku70/Ku80. Here, we identified another function of PHF1 as a potential p53 pathway activator in a pathway screen using luminescence reporter assay. Subsequent studies showed PHF1 directly interacts with p53 proteins both in vivo and in vitro and co-localized in nucleus. PHF1 binds to the C-terminal regulatory domain of p53. Overexpression of PHF1 elevated p53 protein level and prolonged its turnover. Knockdown of PHF1 reduced p53 protein level and its target gene expression both in normal state and DNA damage response. Mechanically, PHF1 protects p53 proteins from MDM2-mediated ubiquitination and degradation. Furthermore, we showed that PHF1 regulates cell growth arrest and etoposide-induced apoptosis in a p53-dependent manner. Finally, PHF1 expression was significantly down-regulated in human breast cancer samples. Taken together, we establish PHF1 as a novel positive regulator of the p53 pathway. These data shed light on the potential roles of PHF1 in tumorigenesis and/or tumor progression. PMID:23150668

  17. Responses of Crop Pests and Natural Enemies to Wildflower Borders Depends on Functional Group.

    Science.gov (United States)

    McCabe, Ellie; Loeb, Gregory; Grab, Heather

    2017-07-25

    Increased homogeneity of agricultural landscapes in the last century has led to a loss of biodiversity and ecosystem services. However, management practices such as wildflower borders offer supplementary resources to many beneficial arthropods. There is evidence that these borders can increase beneficial arthropod abundance, including natural enemies of many pests. However, this increase in local habitat diversity can also have effects on pest populations, and these effects are not well-studied. In this study, we investigated how wildflower borders affect both natural enemies and pests within an adjacent strawberry crop. Significantly more predators were captured in strawberry plantings with wildflower borders versus plantings without wildflowers, but this effect depended on sampling method. Overall, herbivore populations were lower in plots with a wildflower border; however, responses to wildflower borders varied across specific pest groups. Densities of Lygus lineolaris (Tarnished Plant Bug), a generalist pest, increased significantly in plots that had a border, while Stelidota geminata (Strawberry Sap Beetle) decreased in strawberry fields with a wildflower border. These results suggest that wildflower borders may support the control of some pest insects; however, if the pest is a generalist and can utilize the resources of the wildflower patch, their populations may increase within the crop.

  18. Effect of error propagation of nuclide number densities on Monte Carlo burn-up calculations

    International Nuclear Information System (INIS)

    Tohjoh, Masayuki; Endo, Tomohiro; Watanabe, Masato; Yamamoto, Akio

    2006-01-01

    As a result of improvements in computer technology, the continuous energy Monte Carlo burn-up calculation has received attention as a good candidate for an assembly calculation method. However, the results of Monte Carlo calculations contain the statistical errors. The results of Monte Carlo burn-up calculations, in particular, include propagated statistical errors through the variance of the nuclide number densities. Therefore, if statistical error alone is evaluated, the errors in Monte Carlo burn-up calculations may be underestimated. To make clear this effect of error propagation on Monte Carlo burn-up calculations, we here proposed an equation that can predict the variance of nuclide number densities after burn-up calculations, and we verified this equation using enormous numbers of the Monte Carlo burn-up calculations by changing only the initial random numbers. We also verified the effect of the number of burn-up calculation points on Monte Carlo burn-up calculations. From these verifications, we estimated the errors in Monte Carlo burn-up calculations including both statistical and propagated errors. Finally, we made clear the effects of error propagation on Monte Carlo burn-up calculations by comparing statistical errors alone versus both statistical and propagated errors. The results revealed that the effects of error propagation on the Monte Carlo burn-up calculations of 8 x 8 BWR fuel assembly are low up to 60 GWd/t

  19. An economic evaluation of a storage system for casks with burnup credit

    International Nuclear Information System (INIS)

    Mimura, Masahiro; Tsuda, Kazuaki; Yamada, Nobuyuki; O-iwa, Akio.

    1993-01-01

    It is generally recognized that casks designed with burnup credit are more economical than those without burnup credit. To estimate how much more economical they are, we made conceptual designs of transport/storage casks with and without burnup credit for PWR and BWR fuels of various uranium enrichment. The casks were designed to contain the maximum number of fuel assemblies under the necessary weight and dimensional limitations as well as the criticality and shielding criteria. The results showed that approximately 8 % to 44 % more fuel assemblies could be contained in casks with burnup credit. We then evaluated the economy of cask storage system incorporating the cask designs obtained above both with and without burnup credit. The results showed that the cost of storing casks with burnup credit is approximately 7 % to 30 % less expensive than storing casks without burnup credit. (J.P.N.)

  20. Fission gas and iodine release measured in IFA-430 up to 15 GWd/t UO2 burnup

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Turnbull, J.A.; White, R.J.

    1983-01-01

    The release of fission products from fuel pellets to the fuel-cladding gap is dependent on the fuel temperature, the power (fission rate) and the burnup (fuel structure). As part of the US Nuclear Regulatory Commission's Fuel Behavior Program, EG and G Idaho, Inc., is conducting fission product release studies in the Heavy Boiling Water Reactor in Halden, Norway. This paper presents a summary of the results up to December, 1982. The data cover fuel centerline temperatures ranging from 700 to 1500 0 C for average linear heat ratings of 16 to 35 kW/m. The measurements have been performed for the period between 4.2 and 14.8 GWd/t UO 2 of burnup of the Instrumented Fuel Assembly 430 (IFA-430). The measurement program has been directed toward quantifying the release of the short-lived radioactive noble gases and iodines

  1. Power ramp tests of high burnup BWR segment rods

    International Nuclear Information System (INIS)

    Hayashi, H.; Etoh, Y.; Tsukuda, Y.; Shimada, S.; Sakurai, H.

    2002-01-01

    Lead use assemblies (LUAs) of high burnup 8x8 fuel design for Japanese BWRs were irradiated up to 5 cycles in Fukushima Daini Nuclear Power Station No. 2 Unit. Segment rods were installed in LUAs and used for power ramp tests in Japanese Material Test Reactor (JMTR). Post irradiation examinations (PIEs) of segment rods were carried out at Nippon Nuclear Fuel Development Co., Ltd. before and after ramp tests. Maximum linear heat rates of LUAs were kept above 300 W/cm in the first cycle, above 250 W/cm in the second and third cycles and decreased to 200 W/cm in the fourth cycle and 80 W/cm in the fifth cycle. The integrity of high burnup 8x8 fuel was confirmed up to the bundle burnup of 48 GWd/t after 5 cycles of irradiation. Systematic and high quality data were collected through detailed PIEs. The main results are as follows. The oxide on the outer surface of cladding tubes was uniform and its thickness was less than 20 micro-meter after 5 cycles of irradiation and was almost independent of burnup. Hydrogen contents in cladding tubes were less than 150 ppm after 5 cycles of irradiation, although hydrogen contents increased during the fourth and fifth irradiation cycles. Mechanical properties of cladding tubes were on the extrapolated line of previous data up to 5 cycles of irradiation. Fission gas release rates were in the low level (mainly less than 6%) up to 5 cycles of irradiation due to the design to decrease pellet temperature. Pellet-cladding bonding layers were observed after the third cycle and almost full bonding was observed after the fifth cycle. Pellet volume increased with burnup in proportion to solid swelling rate up to the forth cycle. After the fifth cycle, slightly higher pellet swelling was confirmed. Power ramp tests were carried out and satisfactory performance of Zr-lined cladding tube was confirmed up to 60 GWd/t (segment average burnup). One segment rod irradiated for 3 cycles failed by a single step ramp test at terminal ramp power of 614 W

  2. The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu

    2000-07-01

    The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)

  3. Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

    International Nuclear Information System (INIS)

    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

  4. UABUC - Single energy point model burnup computer code for water reactors

    International Nuclear Information System (INIS)

    El-Meshad, Y.; Morsy, S.; El-Osery, I.A.

    1981-01-01

    UABUC is a single energy point reactor burnup computer program in FORTRAN language. The program calculates the change in the isotopic composition of the uranium fuel as a function of irradiation time with all its associated quantities such as the average point flux, the conversion ratio, macroscopic fuel cross sections, and the point reactivity profile. A step-wise time analytical solution was developed for the nonlinear first order burnup differential equations. The ''Westcott'' convention of the effective cross sections was used except for plutonium-240 and uranium-238. For plutonium-240, an effective microscopic cross section was derived from the direct physical arguments taking into account the selfshielding effect of plutonium-240 as well as the 1 ev. resonance absorption. For uranium-238, an effective cross section, reflecting the effect of fast fission and resonance absorption was used. The fission products were treated in the three groups with 50, 300, and 800 barns. The yields in the groups were treated as functions of the type of fissionable nuclides, the effective neutron temperature, and the epithermal index. Xenon-135 and Samarium-149 were treated separately as functions of irradiation time. (author)

  5. The estimation of the control rods absorber burn-up during the VVER-1000 operation

    Energy Technology Data Exchange (ETDEWEB)

    Bolshagin, Sergey N.; Gorodkov, Sergey S.; Sukhino-Khomenko, Evgeniya A. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2013-09-15

    The isotopic composition of the control rods absorber changes under the neutron flux influence, so the control rods efficiency can decrease. In the VVER-1000 control rods boron carbide and dysprosium titanate are used as absorbing materials. In boric part the efficiency decreases due to the {sup 10}B isotope burn-up. Dysprosium isotopes turn into other absorbing isotopes, so the absorbing properties of dysprosium part decrease to a lesser degree. Also the control rod's shells may be deformed as a consequence of boron carbide radiation swelling. This fact should be considered in substantiation of control rods durability. For the estimation of the control rods absorber burn-up two models are developed: VVER-1000 3-D fuel assembly with control rods partially immersed (imitation of the control rods operation in the working group) and VVER-1000 3-D fuel assembly with control rods, located at the upper limit switch (imitation of the control rods operation in groups of the emergency shutdown system). (orig.)

  6. Protein farnesyltransferase isoprenoid substrate discrimination is dependent on isoprene double bonds and branched methyl groups.

    Science.gov (United States)

    Micali, E; Chehade, K A; Isaacs, R J; Andres, D A; Spielmann, H P

    2001-10-16

    Farnesylation is a posttranslational lipid modification in which a 15-carbon farnesyl isoprenoid is linked via a thioether bond to specific cysteine residues of proteins in a reaction catalyzed by protein farnesyltransferase (FTase). We synthesized the benzyloxyisoprenyl pyrophosphate (BnPP) series of transferable farnesyl pyrophosphate (FPP) analogues (1a-e) to test the length dependence of the isoprenoid substrate on the FTase-catalyzed transfer of lipid to protein substrate. Kinetic analyses show that pyrophosphates 1a-e and geranyl pyrophosphate (GPP) transfer with a lower efficiency than FPP whereas geranylgeranyl pyrophosphate (GGPP) does not transfer at all. While a correlation was found between K(m) and analogue hydrophobicity and length, there was no correlation between k(cat) and these properties. Potential binding geometries of FPP, GPP, GGPP, and analogues 1a-e were examined by modeling the molecules into the active site of the FTase crystal structure. We found that analogue 1d displaces approximately the same volume of the active site as does FPP, whereas GPP and analogues 1a-c occupy lesser volumes and 1e occupies a slightly larger volume. Modeling also indicated that GGPP adopts a different conformation than the farnesyl chain of FPP, partially occluding the space occupied by the Ca(1)a(2)X peptide in the ternary X-ray crystal structure. Within the confines of the FTase pocket, the double bonds and branched methyl groups of the geranylgeranyl chain significantly restrict the number of possible conformations relative to the more flexible lipid chain of analogues 1a-e. The modeling results also provide a molecular explanation for the observation that an aromatic ring is a good isostere for the terminal isoprene of FPP.

  7. Potentiation of NMDA receptor-dependent cell responses by extracellular high mobility group box 1 protein.

    Directory of Open Access Journals (Sweden)

    Marco Pedrazzi

    Full Text Available BACKGROUND: Extracellular high mobility group box 1 (HMGB1 protein can operate in a synergistic fashion with different signal molecules promoting an increase of cell Ca(2+ influx. However, the mechanisms responsible for this effect of HMGB1 are still unknown. PRINCIPAL FINDINGS: Here we demonstrate that, at concentrations of agonist per se ineffective, HMGB1 potentiates the activation of the ionotropic glutamate N-methyl-D-aspartate receptor (NMDAR in isolated hippocampal nerve terminals and in a neuroblastoma cell line. This effect was abolished by the NMDA channel blocker MK-801. The HMGB1-facilitated NMDAR opening was followed by activation of the Ca(2+-dependent enzymes calpain and nitric oxide synthase in neuroblastoma cells, resulting in an increased production of NO, a consequent enhanced cell motility, and onset of morphological differentiation. We have also identified NMDAR as the mediator of HMGB1-stimulated murine erythroleukemia cell differentiation, induced by hexamethylenebisacetamide. The potentiation of NMDAR activation involved a peptide of HMGB1 located in the B box at the amino acids 130-139. This HMGB1 fragment did not overlap with binding sites for other cell surface receptors of HMGB1, such as the advanced glycation end products or the Toll-like receptor 4. Moreover, in a competition assay, the HMGB1((130-139 peptide displaced the NMDAR/HMGB1 interaction, suggesting that it comprised the molecular and functional site of HMGB1 regulating the NMDA receptor complex. CONCLUSION: We propose that the multifunctional cytokine-like molecule HMGB1 released by activated, stressed, and damaged or necrotic cells can facilitate NMDAR-mediated cell responses, both in the central nervous system and in peripheral tissues, independently of other known cell surface receptors for HMGB1.

  8. Burnup effects on criticality, breeding and safety of 1,000 MWe gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Ohta, Fumio

    1977-12-01

    Burnup characteristics of 1,000 MWe, PuO 2 - UO 2 fuelled helium-cooled fast breeder reactor have been studied concerning criticality, breeding and safety. A 26-energy group cross-section set produced from ENDF/B-3 was used. Criticality and breeding were studied with two-dimensional burnup code APOLLO and 4-energy group cross-section set generated by collapsing the mentioned cross-section set. Safety aspects such as Doppler reactivity effect, coolant-depressurisation and steam-ingression reactivity effect were studied with multi-dimensional diffusion theory code CITATION and perturbation theory code PERKY, as well as the 26-energy group cross-section set. The following were revealed: (1) The reactivity swing over a year's irradiation is merely 1.5% ΔK/K. This small swing may permit relatively long fuel dwelling in GCFR and , thus, the frequency of outages for refuelling can be minimised. (2) The surplus fissile plutonium over a year's irradiation is about 360 Kg, and the system doubling time is about 9 years. The GCFR studied has excellent breeding, compared with those in PuO 2 -UO 2 fuelled LMFBR and other GCFRs. (3) The coolant-depressurisation reactivity effect becomes more positive with burnup. This is not so serious as the sodium-void reactivity effect of LMFBR. (4) In the start-up core, the steam-ingression reactivity effect due to steam ingression to the core and blanket from the secondary coolant system becomes positive at certain steam density (0.02gr/cc) and this positive effect increases with steam density. With advance of burnup, however, the effect becomes negative, this increasing with steam density. After all, the steam ingression is no hazard in operation of GCFR since the reactivity effect is negative in the equilibrium state. (auth.)

  9. Modeling of burnup express-estimation for UO{sub 2}-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Likhanskii, Vladimir V.; Tokarev, Sergey A.; Vilkhivskaya, Olga V., E-mail: vilhivskaya_olga@mail.ru

    2017-03-15

    Highlights: • Proposed engineering model estimates fuel burnup by {sup 134}Cs/{sup 137}Cs activity ratio. • Buildup of cesium isotopes relies on changing neutron spectrum in the core cycle. • {sup 134}Cs/{sup 137}Cs activity ratios in FAs with Gd-doped fuel rods are analyzed. • Comparison of the model calculations with the NPPs spike measurements is presented. - Abstract: The paper presents the developed engineering model of cesium isotopes production as function of UO{sub 2}-fuel burnup and an assessment of their activity ratios. The model considers the evolution of linear power of gadolinium-doped fuel rods and fuel rods surrounding them in fuel assemblies with high enrichment fuel, harder neutron spectrum, and the changes in cross-sections of neutron reactions in thermal and epithermal energy areas. Parametrical dependences in the model are based on the fuel operation data for nuclear power plants and on the detailed neutronic-physical calculations of the core. Presented are the results of the model calculations for the {sup 134}Cs/{sup 137}Cs activity ratios in fuel taking into account the parameter of hardness of the neutron spectrum during the first irradiation cycle for fuel with enrichment ranging from 3.6 wt% in {sup 235}U.

  10. Multiplex ligation-dependent probe amplification (MLPA) assay for blood group genotyping, copy number quantification, and analysis of RH variants

    NARCIS (Netherlands)

    Veldhuisen, Barbera; van der Schoot, C. E.; de Haas, Masja

    2015-01-01

    The blood group multiplex ligation-dependent probe amplification (MLPA) is a comprehensive assay, developed for genotyping the majority of clinically relevant blood group antigens in both patients and donors. The MLPA is an easy method to apply and only requires a thermal cycler and capillary

  11. Exploring Secondary Students' Epistemological Features Depending on the Evaluation Levels of the Group Model on Blood Circulation

    Science.gov (United States)

    Lee, Shinyoung; Kim, Heui-Baik

    2014-01-01

    The purpose of this study is to identify the epistemological features and model qualities depending on model evaluation levels and to explore the reasoning process behind high-level evaluation through small group interaction about blood circulation. Nine groups of three to four students in the eighth grade participated in the modeling practice.…

  12. 75 FR 27141 - Group Health Plans and Health Insurance Issuers Providing Dependent Coverage of Children to Age...

    Science.gov (United States)

    2010-05-13

    ... Group Health Plans and Health Insurance Issuers Providing Dependent Coverage of Children to Age 26 Under... Information and Insurance Oversight of the U.S. Department of Health and Human Services are issuing substantially similar interim final regulations with respect to group health plans and health insurance coverage...

  13. CARA design criteria for HWR fuel burnup extension

    International Nuclear Information System (INIS)

    Florido, P.C.; Cirimello, R.O.; Bergallo, J.E.; Marino, A.C.; Delmastro, D.F.; Brasnarof, D.O.; Gonzalez, J.H.; Juanico, L.A.

    2002-01-01

    A new concept for HWR fuel bundles, namely CARA, is presented. The CARA design allows to improve all the major performances in the PHWR fuel technology. Among others, it reaches higher burnup and thermohydraulic safety margins, together with lower fuel pellet temperatures and Zry/HM mass ratio. Moreover, it keeps the fuel mass content per unit length and the channel pressure drop by using a single diameter of fuel rods. (author)

  14. Effect of fuel burnup on the mechanical safety coefficients

    International Nuclear Information System (INIS)

    Plyashkevich, V.Ju.; Sidorenko, V.D.; Shishkov, L.K.

    2001-01-01

    )In the paper the results of studies of changes in the process of campaign 'disturbances' of local heat flux and local fuel burnup, resulting from the 'mechanical' deviations in the composition and geometrical characteristics of fuel rods from the nominal are given. As example, the WWER-440 fuel assembly with burnable poisons used in the five-year fuel cycle is considered. The effect of deviations in fuel enrichment, fuel content, gadolinium content and geometrical size was studied (Authors)

  15. Validation of SCALE-4 for burnup credit applications

    International Nuclear Information System (INIS)

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

    1995-01-01

    In the past, a criticality analysis of PWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at ORNL in support of the US DOE efforts to demonstrate a validation approach for criticality safety methods to be used in burnup credit cask design. To date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The ANSI/ANS-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors by correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k eff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications

  16. Calculation of triton confinement and burn-up in tokamaks

    International Nuclear Information System (INIS)

    Anderson, D.; Battistoni, P.

    1987-01-01

    An analytical investigation is made of the confinement and subsequent burn-up of fusion produced tritons in a deuterium Tokamak plasma. Explicit approximations are obtained for the triton confinement factor, clearly displaying the scaling with physical parameters. The importance of pitch angle scattering losses during the triton slowing down is also estimated. A comparison with experiments and numerical calculations on the FT Tokamak slows good qualitative agreement. (authors)

  17. Investigation of Burnup Credit Issues in BWR Fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.; DeHart, M.D.

    1999-01-01

    Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of the large mass of fissile material that will be present in the repository. Burnup credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents, followed by criticality calculations to assess the value of keff for a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel

  18. Three dimensional Burn-up program parallelization using socket programming

    International Nuclear Information System (INIS)

    Haliyati R, Evi; Su'ud, Zaki

    2002-01-01

    A computer parallelization process was built with a purpose to decrease execution time of a physics program. In this case, a multi computer system was built to be used to analyze burn-up process of a nuclear reactor. This multi computer system was design need using a protocol communication among sockets, i.e. TCP/IP. This system consists of computer as a server and the rest as clients. The server has a main control to all its clients. The server also divides the reactor core geometrically to in parts in accordance with the number of clients, each computer including the server has a task to conduct burn-up analysis of 1/n part of the total reactor core measure. This burn-up analysis was conducted simultaneously and in a parallel way by all computers, so a faster program execution time was achieved close to 1/n times that of one computer. Then an analysis was carried out and states that in order to calculate the density of atoms in a reactor of 91 cm x 91 cm x 116 cm, the usage of a parallel system of 2 computers has the highest efficiency

  19. Observations on the CANDLE burn-up in various geometries

    International Nuclear Information System (INIS)

    Seifritz, W.

    2007-01-01

    We have looked at all geometrical conditions under which an auto catalytically propagating burnup wave (CANDLE burn-up) is possible. Thereby, the Sine Gordon equation finds a new place in the burn-up theory of nuclear fission reactors. For a practical reactor design the axially burning 'spaghetti' reactor and the azimuthally burning 'pancake' reactor, respectively, seem to be the most promising geometries for a practical reactor design. Radial and spherical burn-waves in cylindrical and spherical geometry, respectively, are principally impossible. Also, the possible applicability of such fission burn-waves on the OKLO-phenomenon and the GEOREACTOR in the center of Earth, postulated by Herndon, is discussed. A fast CANDLE-reactor can work with only depleted uranium. Therefore, uranium mining and uranium-enrichment are not necessary anymore. Furthermore, it is also possible to dispense with reprocessing because the uranium utilization factor is as high as about 40%. Thus, this completely new reactor type can open a new era of reactor technology

  20. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

    2003-01-01

    The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

  1. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

    2015-01-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  2. CEA contribution to power plant operation with high burnup level

    International Nuclear Information System (INIS)

    1981-03-01

    High level burnup in PWR leads to investigate again the choices carried out in the field of fuel management. French CEA has studied the economic importance of reshuffling technique, cycle length, discharge burnup, and non-operation period between two cycles. Power plants operators wish to work with increased length cycles of 18 months instead of 12. That leads to control problems because the core reactivity cannot be controlled with the only soluble boron: moderator temperature coefficient must be negative. With such cycles, it is necessary to use burnable poisons and for economic reasons with a low penalty in end of cycle. CEA has studied the use of Gd 2 O 3 mixed with fuel or with inert element like Al 2 O 3 . Parametric studies of specific weights, efficacities relatively to the fuel burnup and the fuel enrichment have been carried out. Particular studies of 1 month cycles with Gd 2 O 3 have shown the possibility to control power distribution with a very low reactivity penalty in EOC. In the same time, in the 100 MW PWR-CAP, control reactivity has been made with large use of gadolinia in parallel with soluble boron for the two first cycles

  3. Advanced fuel cycles and burnup increase of WWER-440 fuel

    International Nuclear Information System (INIS)

    Proselkov, V.; Saprykin, V.; Scheglov, A.

    2003-01-01

    Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

  4. Advances in fuel pellet technology for improved performance at high burnup. Proceedings of a Technical Committee meeting

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA has recently completed two co-ordinated Research Programmes (CRPs) on The Development of Computer Models for Fuel Element Behaviour in Water Reactors, and on Fuel Modelling at Extended Burnup. Through these CRPs it became evident that there was a need to obtain data on fuel behaviour at high burnup. Data related o thermal behaviour, fission gas release and pellet to clad mechanical interaction were obtained and presented at the Technical Committee Meeting on Advances in Fuel Pellet Technology for Improved Performance at High Burnup which was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). The 34 papers from 10 countries are published in this proceedings and presented by a separate abstract. The papers were grouped in 6 sessions. First two sessions covered the fabrication of both UO 2 fuel and additives and MOX fuel. Sessions 3 and 4 covered the thermal behaviour of both types of fuel. The remaining two sessions dealt with fission gas release and the mechanical aspects of pellet to clad interaction

  5. Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

    2009-01-01

    In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

  6. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  7. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    International Nuclear Information System (INIS)

    Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

    2014-01-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

  8. KENOREST - A new coupled code system based on KENO and OREST for criticality and burnup inventory calculations

    International Nuclear Information System (INIS)

    Hesse, U.; Gmal, B.; Voggenberger, Th.; Baleanu, M.; Langenbuch, S.

    2001-01-01

    The program system KENOREST version 1998 will be presented, which is a useful tool for burnup and reactivity calculations for LWR fuel. The three-dimensional Monte Carlo code KENO-V.a is coupled with the one-dimensional GRS burnup program system OREST-98. The objective is to achieve a better modelling of plutonium and actinide build-up or burnout for advanced heterogeneous fuel assembly designs. Further objectives are directed to reliable calculations of the pin power distributions and of reactor safety parameters including axial and radial rod temperatures for fuel assemblies of modern design. The stand-alone-code KENO-V.a version is used without any changes in the program source. The OREST-98 system was developed to handle multirod problems and additional burnup dependent moderator conditions which can be applied to stretch-out simulations in the reactor. A new interface module RESPEFF between KENO and OREST transforms the 2-d or 3-d KENO flux results to the one-dimensional lattice code OREST in a fully automated manner to maintain reaction rate balance between the codes. First results for assembly multiplication factors, isotope inventories are compared with OECD results. (author)

  9. Numeric modeling of HfO2 neutron flux sensor parameters during sensor burnup in the RBMK-1500 reactor

    International Nuclear Information System (INIS)

    Jurkevicius, A.; Remeikis, V.

    2001-01-01

    The isotopic composition of hafnium in the radial neutron flux sensor of the RBMK-1500 reactor, the rates of the neutron absorption on Hf isotopes and the neutron spectrum in the sensor were numerically modeled. The sequence SAS2 (Shielding Analysis Sequence) from the package SCALE 4.3 was used for calculations. It has been obtained that the main neutron absorber 167 Er isotope practically burns up completely at the 18 MW d/kgU burnup depth, and at that time the capture rate of thermal neutrons in erbium decreases ten-fold. The average neutron flux density was calculated 7.6*10 13 neutrons. Cm -2 S -1 in the RBMK-1500 reactor grating, when the nuclear fuel enriched with 235 U by 2.4% and with Er by 0.4% is used in a fuel assembly. When the sensor burnup reaches 28 MW d/kgU, the neutron absorption rate of 178 Hf exceeds the rate of 177 Hf. The overall neutron absorption rate in hafnium decreases 2.53 times due to the sensor burnup to 56 MW d/kgU. The corrective factors ξ d (I) at different integral flux I of the sensors were calculated. The obtained dependence ξ d (I) calculated numerically was compared to the experimental one determined by processing repeated calibration results of Hf sensors in RBMK-1500 reactors, as well as compared to the theoretical one currently used in the Ignalina NPP special mathematical algorithms. (author)

  10. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  11. THE DIFFERENCES IN MORAL, GROUP IDENTITY AND THE PERCON’S VARIABILITY DEPENDING ON THE EDUCATION

    Directory of Open Access Journals (Sweden)

    Irina Aleksandrobna Kolinichenko

    2017-06-01

    Results. The results of the study have revealed the dominance of all specified assessment parameters in the group of test subjects with incomplete higher education: higher level of moral development in all dilemmas (the opposition of life values (compassion and following the law, self-interest – the interests of the city (law, business (benefit and law, personal interests (career and the freedom of another person, except for the dilemma of the opposition between the interests of a majority and a single person. The differences have also been revealed between the two groups of test subjects according to the group identity, group variability, the desirability of the common categories of identity.

  12. The Effect of Group Therapy With Transactional Analysis Approach on Emotional Intelligence, Executive Functions and Drug Dependency.

    Science.gov (United States)

    Forghani, Masoomeh; Ghanbari Hashem Abadi, Bahram Ali

    2016-06-01

    The aim of the present study was to evaluate the effect of group psychotherapy with transactional analysis (TA) approach on emotional intelligence (EI), executive functions and substance dependency among drug-addicts at rehabilitation centers in Mashhad city, Iran, in 2013. In this quasi-experimental study with pretest, posttest, case- control stages, 30 patients were selected from a rehabilitation center and randomly divided into two groups. The case group received 12 sessions of group psychotherapy with transactional analysis approach. Then the effects of independent variable (group psychotherapy with TA approach) on EI, executive function and drug dependency were assessed. The Bar-on test was used for EI, Stroop test for measuring executive function and morphine test, meth-amphetamines and B2 test for evaluating drug dependency. Data were analyzed using multifactorial covariance analysis, Levenes' analysis, MANCOVA, t-student and Pearson correlation coefficient tests t with SPSS software. Our results showed that group psychotherapy with the TA approach was effective in improving EI, executive functions and decreasing drug dependency (P addicts and prevents addiction recurrence by improving the coping capabilities and some mental functions of the subjects. However, there are some limitations regarding this study including follow-up duration and sample size.

  13. Analytic parameter dependence of Harish-Chandra modules for real reductive Lie groups - a family affair

    NARCIS (Netherlands)

    van der Noort, V.

    2009-01-01

    This thesis is written in the subfield of mathematics known as representation theory of real reductive Lie groups. Let G be a Lie group in the Harish-Chandra class with maximal compact subgroup K and Lie algebra g. Let Omega be a connected complex manifold. By a family of G-representations

  14. The effect of ν-bar(E) energy dependence at E 239Pu in the lower energy groups

    International Nuclear Information System (INIS)

    Gusejnov, A.G.; Gusejnov, M.A.; Rabotnov, N.S.

    1994-01-01

    The effect of the energy dependence of the average number of prompt neutrons on the 239 Pu fission BNAB-26 group cross-sections has been estimated. It was also found that the values of ν-bar(E) are lowered by approximately 1% when the temperature increases from 300 deg. K to 2000 deg. K. Taking this dependence into account may improve the predictability of the plutonium-fueled reactor characteristics. (author). 7 refs, 1 fig., 2 tabs

  15. Fuel element burnup determination in HEU-LEU mixed TRIGA research reactor core

    International Nuclear Information System (INIS)

    Zagar, Tomaz; Ravnik, Matjaz

    2000-01-01

    This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

  16. Is in-group bias culture-dependent? A meta-analysis across 18 societies.

    Science.gov (United States)

    Fischer, Ronald; Derham, Crysta

    2016-01-01

    We report a meta-analysis on the relationship between in-group bias and culture. Our focus is on whether broad macro-contextual variables influence the extent to which individuals favour their in-group. Data from 21,266 participants from 18 societies included in experimental and survey studies were available. Using Hofstede's (1980) and Schwartz (2006) culture-level predictors in a 3-level mixed-effects meta-analysis, we found strong support for the uncertainty-reduction hypothesis. An interaction between Autonomy and real vs artificial groups suggested that in low autonomy contexts, individuals show greater in-group bias for real groups. Implications for social identity theory and intergroup conflict are outlined.

  17. On the influence of spatial discretization in LWR steady state and burnup calculations with HELIOS 1.9

    International Nuclear Information System (INIS)

    Merk, B.; Weiss, F. P.

    2009-01-01

    Cell and burnup calculations are fundamental to all deterministic static and transient 3D full core calculations for different operational states of the reactor. The spatial discretization used for the cell and burnup calculations influences significantly the results of full integral transport solutions. The influence of the discretization on k inf is shown for the steady state case and the influence on the neutron spectrum is analyzed. Moreover, the differences in k inf are presented for different spatial discretization strategies in the burnup calculation of Uranium Oxide (UOX) fuel. The resulting different flux distributions cause significant changes in the isotopic densities. The influence of the discretization strategies on the calculation of homogenized few group cross-sections is investigated. This detailed discretization study demonstrates the need for sufficiently fine discretization to produce reliable and accurate results when using integral transport methods. In contrast to the currently used discretization schemes, refined discretization is especially important in the moderator region of the unit cell to reproduce the influence on the thermal neutron spectrum. Additionally, the need for sufficient discretization affects the idea of full core calculations based on integral transport methods since it has to be discussed whether it is worth to do full core calculations with reduced discretization when facing this strong discretization effect. The computer resources required for full core calculations with fine discretization are currently not available. (authors)

  18. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  19. DRAGON, Reactor Cell Calculation System with Burnup

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: DRAGON is a collection of models to simulate the neutronic behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations which can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. The user must supply cross sections. DRAGON can access directly standard microscopic cross-section libraries in the following formats: DRAGON, MATXS (TRANSX-CTR), WIMSD4, WIMS-AECL, and APOLLO. It has the capability of exchanging macroscopic and microscopic cross-section libraries with a code such as PSR-0206/TRANSX-CTR or PSR-0317/TRANSX-2 by the use of the GOXS and ISOTXS format files. Macroscopic cross sections can also be read in DRAGON via the input data stream. 2 - Method of solution: DRAGON contains a multigroup iterator conceived to control a number of different algorithms for the solution of the neutron transport equation. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are included in a source term. The current version, DRAGON 9 71124 (Release 3.02), which was released in January 1998, contains three such algorithms. The JPM option solves the integral transport equation using the interface current method applied to homogeneous blocks; the SYBIL option solves the integral transport equation using the collision probability method for simple one-dimensional (1-D) or two-dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; and the

  20. Effectiveness of Positive Group Psychotherapy on Resiliency, Happiness and General Health on Women with a Substance Dependence Spouses

    Directory of Open Access Journals (Sweden)

    F sohrabi

    2016-03-01

    Full Text Available The aim of this study was to evaluate the effectiveness of positive group psychotherapy on resiliency, happiness and general health in Women with substance dependent spouse. The research design was quasi - experimental with pre-test, post-test and one-month follow - up with the control group. The sample was comprised of women whom their spouse were referred to addiction clinic. Twenty - four women were selected and randomly assigned to experimental and control groups. Both groups were assessed by Davidson Resilience Scale, Peterson happiness directions questionnaires and General Health Questionnaire (GHQ. Positive group psychotherapy was conducted on the experimental group for 14 sessions, each session lasting 90- minutes. The data was analyzed with ANCOVA. The findings indicated that the mean scores of post-test in the experimental groups on resilience, happiness, and public health, was significantly higher than the mean scores of post-test in control group. Benferroni test was indicating the stable effect of therapy. According to the findings, it can be said that group positive psychotherapy was effective in promoting resiliency, happiness, and general health of women with the dependent drug spouses.

  1. Polynomial expansion methodology for microscopic cross sections to use in spatial burnup calculations

    International Nuclear Information System (INIS)

    Conti Filho, P.; Oliveira Barroso, A.C. de

    1985-01-01

    It was developed a computer code to generate polynomial coefficients which represent homogenized microscopic cross sections in function of the local accumulated burnup and concentration of soluble boron, presented in fuel element, for each step of burnup reactor. Afterward, it was developed a coupling between LEOPARD-GERADOR DE POLINOMIOS - CITATION computer codes to interpret and build homogenized microscopic cross sections according with local characteristics of each fuel element during the burnup calculation of reactor core. (M.C.K.) [pt

  2. Comparison of measured and calculated burn-up of AVR-Fuel-Elements

    Energy Technology Data Exchange (ETDEWEB)

    Wagemann, R.

    1974-03-15

    Burn-up comparisons are made for small batches of three types of AVR fuel elements using a coupled EREBUS-MUPO neutronic analysis compared against test results from both nondestructive gamma-ray measurements of cesium-137 activity and destructive mass spectrometry measurements of the ratio of U-233 to U-235. The comparisons are relatively good for average burn-up and reasonably good for burn-up distributions.

  3. Effect of local burn-up variation on computed mean nuclide concentrations

    International Nuclear Information System (INIS)

    Moeller, W.

    1982-01-01

    Mean concentrations of U-235, U-236, U-238, Pu-239, Pu-240, Pu-241 and Pu-242 in some volume areas of WWER-440 fuel assemblies have been calculated from corresponding burn-up microdistribution data and compared with those calculated from burn-up mean values. Differences occurring were below 3% for the uranium nuclides but, at low burn-ups, considerable for Pu-241 and Pu-242. (author)

  4. Functional delineation of three groups of the ATP-dependent family of chromatin remodeling enzymes.

    NARCIS (Netherlands)

    Boyer, L.A.; Logie, C.; Bonte, E; Becker, P.B.; Wade, P.A.; Wolff, A.P.; Wu, C.; Imbalzano, A.N.; Peterson, C.L.

    2000-01-01

    ATP-dependent chromatin remodeling enzymes antagonize the inhibitory effects of chromatin. We compare six different remodeling complexes: ySWI/SNF, yRSC, hSWI/SNF, xMi-2, dCHRAC, and dNURF. We find that each complex uses similar amounts of ATP to remodel nucleosomal arrays at nearly identical rates.

  5. Handling requirements dependencies in agile projects: A focus group with agile software development practitioners

    NARCIS (Netherlands)

    Martakis, Aias; Daneva, Maia; Wieringa, Roelf J.; Jean-Louis Cavarero, S.; Rolland, C; Cavarero, J.L.

    2013-01-01

    Agile practices on requirements dependencies are a relatively unexplored topic in literature. Empirical studies on it are scarce. This research sets out to uncover concepts that practitioners in companies of various sizes across the globe and in various industries, use for dealing with requirements

  6. Anomalous length dependence of conductance of aromatic nanoribbons with amine anchoring groups

    KAUST Repository

    Bilić, Ante; Sanvito, Stefano

    2012-01-01

    for longer members of the series. The oligoperylene nanoribbons, with dual amine groups at both terminals, show the potential to fully harness the highly conjugated system of π molecular orbitals across the junction. © 2012 American Physical Society.

  7. Calculation of effect of burnup history on spent fuel reactivity based on CASMO5

    International Nuclear Information System (INIS)

    Li Xiaobo; Xia Zhaodong; Zhu Qingfu

    2015-01-01

    Based on the burnup credit of actinides + fission products (APU-2) which are usually considered in spent fuel package, the effect of power density and operating history on k_∞ was studied. All the burnup calculations are based on the two-dimensional fuel assembly burnup program CASMO5. The results show that taking the core average power density of specified power plus a bounding margin of 0.0023 to k_∞, and taking the operating history of specified power without shutdown during cycle and between cycles plus a bounding margin of 0.0045 to k_∞ can meet the bounding principle of burnup credit. (authors)

  8. Nuclear-data uncertainty propagations in burnup calculation for the PWR assembly

    International Nuclear Information System (INIS)

    Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei

    2017-01-01

    Highlights: • The DRAGON 5.0 and NECP-CACTI have been implemented in UNICORN. • The effects of different neutronics methods on S&U results were quantified. • Uncertainty analysis has been applied to burnup calculation of PWR assembly. • The uncertainties of eigenvalue and few-group constants have been quantified. - Abstract: In this paper, our home-developed lattice code NECP-CACTI has been implemented into our UNICORN code to perform sensitivity and uncertainty analysis for the lattice calculations. The verified multigroup cross-section perturbation model and methods of the sensitivity and uncertainty analysis are established and applied to different lattice codes in UNICORN. As DRAGON5.0 and NECP-CACTI are available for the lattice calculations in UNICORN now, the effects of different neutronics methods (including methods for the neutron-transport and resonance self-shielding calculations) on the results of sensitivity and uncertainty analysis were studied in this paper. Based on NECP-CACTI, uncertainty analysis using the statistical sampling method has been performed to the burnup calculation for the fresh-fueled TMI-1 assembly, propagating the nuclear-data uncertainties to k_∞ and two-group constants of the lattice calculation with depletions. As results shown, for different neutronics methods, it can be observed that different methods of the neutron-transport calculation introduce no differences to the results of sensitivity and uncertainty analysis, while different methods of the resonance self-shielding calculation would impact the results. With depletions of the TMI-1 assembly, for k_∞, the relative uncertainty varies between 0.45% and 0.60%; for two-group constants, the largest variation is between 0.35% and 2.56% for vΣ_f_,_2. Moreover, the most significant contributors to the uncertainty of k_∞ and two-group constants varied with depletions are determined.

  9. Escherichia coli Phosphoenolpyruvate-Dependent Phosphotransferase System : Mechanism of Phosphoryl-Group Transfer from Phosphoenolpyruvate to HPr

    NARCIS (Netherlands)

    Misset, Onno; Robillard, George T.

    1982-01-01

    The mechanism of phosphoryl-group transfer from phosphoenolpyruvate (PEP) to HPr, catalyzed by enzyme I of the Escherichia coli PEP-dependent phosphotransferase system, has been studied in vitro. Steady-state kinetics and isotope exchange measurements revealed that this reaction cannot be described

  10. Nuclear fuel behaviour modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2001-07-01

    The Technical Committee Meeting (TCM) included separate sessions on the specific topics of fuel thermal performance and fission product retention. On thermal performance, it is apparent that the capability exists to measure conductivity in high burnup fuel either by out-of-pile measurement or by instrumentation of test reactor rods. State-of-the-art modelling codes contain models for the conductivity degradation process, and hence adequate predictions of fuel temperature are achievable. Concerning fission product release, it is clear that many groups around the world are actively investigating the subject, with experimental and modelling programmes being pursued. However, a general consensus on the exact mechanisms of gas release and related gas bubble swelling has yet to emerge, even at medium burnup levels. Fission gas phenomena, not only the release to open volumes, but the whole sequence of processes taking place prior to this, need to be modelled in any modern fuel performance code. The presence of gaseous fission products may generate rapid fuel swelling during power transients, and this can cause PCI and rod failure. At high burnups, the quantity of released gases could give rise to pressures exceeding the safe limits. Modelling of pellet-cladding interaction (PCI) effects during transient operation is also an active area of study for many groups. In some situations a purely empirical approach to failure modelling can be justified, while for other applications a more detailed mechanistic approach is required. Another aspect of cladding modelling which was featured at the TCM concerned corrosion and hydriding. Although this issue can be the main life-limiting factor on fuel duty, it is apparent that modelling methods, and the experimental measurement techniques that underpin them, are adequate. A session was included on MOX fuel modelling. Substantial programmes of work, especially by the MOX vendors, appear to be underway to bring the level of understanding

  11. PACER: a Monte Carlo time-dependent spectrum program for generating few-group diffusion-theory cross sections

    International Nuclear Information System (INIS)

    Candelore, N.R.; Kerrick, W.E.; Johnson, E.G.; Gast, R.C.; Dei, D.E.; Fields, D.L.

    1982-09-01

    The PACER Monte Carlo program for the CDC-7600 performs fixed source or eigenvalue calculations of spatially dependent neutron spectra in rod-lattice geometries. The neutron flux solution is used to produce few group, flux-weighted cross sections spatially averaged over edit regions. In general, PACER provides environmentally dependent flux-weighted few group microscopic cross sections which can be made time (depletion) dependent. These cross sections can be written in a standard POX output file format. To minimize computer storage requirements, PACER allows separate spectrum and edit options. PACER also calculates an explicit (n, 2n) cross section. The PACER geometry allows multiple rod arrays with axial detail. This report provides details of the neutron kinematics and the input required

  12. Multiplicity dependence of matrix-induced frequency shifts for atomic transitions of the group 12 metals in rare gas solids

    International Nuclear Information System (INIS)

    Laursen, S.L.; Cartland, H.E.

    1991-01-01

    Atomic resonances of the group 12 metal atoms, Hg, Cd, and Zn, undergo frequency shifts from the gas phase atomic line when trapped in rare gas matrices of Ar, Kr, and Xe at 12 K. As expected, the shifts are approximately linear in polarizability of the rare gas, but the slope of this line depends on whether the transition in question is 1 P 1 left-arrow 1 S 0 or 3 P 1 left-arrow 1 S 0 . Thus the matrix-induced frequency shift is dependent on the singlet or triplet nature of the excited state as well as on the matrix material. This dependence on multiplicity is discussed in terms of interactions between the excited-state atomic orbitals and the matrix. The results are compared to matrix studies of other metals and to related gas-phase work on diatomic van der Waals complexes of group 12 metals with rare gases

  13. Calculation of isotope burn-up and change in efficiency of absorbing elements of WWER-1000 control and protection system during burn-up

    International Nuclear Information System (INIS)

    Timofeeva, O.A.; Kurakin, K.U.

    2006-01-01

    The report deals with fast and thermal neutron flows distribution in structural elements of WWER-1000 fuel assembly and absorbing rods, determination of absorbing isotope burn-up and worth variation in WWER reactor control and protection system rods. Simulation of absorber rod burn-up is provided using code package SAPPHIRE 9 5 end RC W WER allowing detailed description of the core segment spatial model. Maximum burn-up of absorbing rods and respective worth variation of control and protection system rods is determined on the basis of a number of calculations considering known characteristics of fuel cycles (Authors)

  14. Changes of the inventory of radioactive materials in reactor fuel from uranium in changing to higher burn-up and determining the important effects of this

    International Nuclear Information System (INIS)

    Kirchner, G.; Schaefer, R.

    1985-01-01

    The knowledge of the nuclide composition during and after use in the reactor is an essential, in order to be able to determine the effects associated with the operation of nuclear plants. The missing reliable data on the inventory of radioactive materials resulting from the expected change to higher burn-ups of uranium fuels in West Germany are calculated. The reliability of the program system used for this, which permits a one-dimensional account taken of the fuel rod cell and measurement of the changes of specific sets of nuclear data depending on burn-up, is confirmed by the comparison with experimentally found concentrations of important nuclides in fuel samples at Obrigheim nuclear power station. Realistic conditions of use are defined for a range of burn-up of 33 GWd/t to 55 GWd/t and the effects of changes of the number of cycles and the use of types of fuel elements being developed on the composition of the inventory are determined. The plutonium compositions during use in the reactor are given and are tabulated with the inventory for decay times up to 30 years. Effects during change to higher burn-ups are examined and discussed for the maximum inventories during use of fuel and for heat generation during final storage. (orig./HP) [de

  15. Differential Effectiveness of Interdependent and Dependent Group Contingencies in Reducing Disruptive Classroom Behavior

    Science.gov (United States)

    Hartman, Kelsey; Gresham, Frank

    2016-01-01

    Disruptive behavior in the classroom negatively affects all students' academic engagement, achievement, and behavior. Group contingencies have been proven effective in reducing disruptive behavior as part of behavior interventions in the classroom. The Good Behavior Game is a Tier 1 classwide intervention that utilizes an interdependent group…

  16. Anomalous length dependence of conductance of aromatic nanoribbons with amine anchoring groups

    KAUST Repository

    Bilić, Ante

    2012-09-06

    Two sets of aromatic nanoribbons, based around a common hexagonal scaffolding, with single and dual terminal amine groups have been considered as potential molecular wires in a junction formed by gold leads. Charge transport through the two-terminal device has been modeled using density functional theory (with and without self-interaction correction) and the nonequilibrium Green\\'s function method. The effects of wire length, multiple terminal contacts, and pathways across the junction have been investigated. For nanoribbons with the oligopyrene motif and conventional single amine terminal groups, an increase in the wire length causes an exponential drop in the conductance. In contrast, for the nanoribbons with the oligoperylene motif and dual amine anchoring groups the predicted conductance rises with the wire length over the whole range of investigated lengths. Only when the effects of self-interaction correction are taken into account, the conductance of the oligoperylene ribbons exhibits saturation for longer members of the series. The oligoperylene nanoribbons, with dual amine groups at both terminals, show the potential to fully harness the highly conjugated system of π molecular orbitals across the junction. © 2012 American Physical Society.

  17. Renormalization group analysis of the temperature dependent coupling constant in massless theory

    International Nuclear Information System (INIS)

    Yamada, Hirofumi.

    1987-06-01

    A general analysis of finite temperature renormalization group equations for massless theories is presented. It is found that in a direction where momenta and temperature are scaled up with their ratio fixed the coupling constant behaves in the same manner as in zero temperature and that asymptotic freedom at short distances is also maintained at finite temperature. (author)

  18. In Peer Matters, Teachers Matter: Peer Group Influences on Students' Engagement Depend on Teacher Involvement

    Science.gov (United States)

    Vollet, Justin W.; Kindermann, Thomas A.; Skinner, Ellen A.

    2017-01-01

    This study focused on the joint effects of teachers and peer groups as predictors of change in students' engagement during the first year of middle school, when the importance of peer relationships normatively increases and the quality of teacher-student relationships typically declines. To explore cumulative and contextualized joint effects, the…

  19. Papiliochrome II pigment reduces the angle dependency of structural wing colouration in nireus group papilionids

    NARCIS (Netherlands)

    Wilts, Bodo D.; Trzeciak, Tomasz M.; Vukusic, Peter; Stavenga, Doekele G.

    The wings of four papilionid butterfly species of the nireus group, Papilio bromius, P. epiphorbas, P. nireus and P. oribazus, are marked by blue-green coloured bands surrounded by black margins. The cover scales in the coloured bands contain a violet-absorbing, blue-fluorescing pigment. The

  20. Group Counselling on College Students' Internet Dependency and Life Satisfaction

    Science.gov (United States)

    Odaci, Hatice; Çelik, Çigdem Berber

    2017-01-01

    The limited number of programs of tested efficacy in the literature such as cognitive-behavioural therapy and family-based prevention of internet addiction is striking. The aim of this study was to analyse the effect of reality therapy-based group counselling on college students' problematic internet use and life satisfaction. In order to…

  1. Holistic face perception is modulated by experience-dependent perceptual grouping.

    Science.gov (United States)

    Curby, Kim M; Entenman, Robert J; Fleming, Justin T

    2016-07-01

    What role do general-purpose, experience-sensitive perceptual mechanisms play in producing characteristic features of face perception? We previously demonstrated that different-colored, misaligned framing backgrounds, designed to disrupt perceptual grouping of face parts appearing upon them, disrupt holistic face perception. In the current experiments, a similar part-judgment task with composite faces was performed: face parts appeared in either misaligned, different-colored rectangles or aligned, same-colored rectangles. To investigate whether experience can shape impacts of perceptual grouping on holistic face perception, a pre-task fostered the perception of either (a) the misaligned, differently colored rectangle frames as parts of a single, multicolored polygon or (b) the aligned, same-colored rectangle frames as a single square shape. Faces appearing in the misaligned, differently colored rectangles were processed more holistically by those in the polygon-, compared with the square-, pre-task group. Holistic effects for faces appearing in aligned, same-colored rectangles showed the opposite pattern. Experiment 2, which included a pre-task condition fostering the perception of the aligned, same-colored frames as pairs of independent rectangles, provided converging evidence that experience can modulate impacts of perceptual grouping on holistic face perception. These results are surprising given the proposed impenetrability of holistic face perception and provide insights into the elusive mechanisms underlying holistic perception.

  2. Development and verification of Monte Carlo burnup calculation system

    International Nuclear Information System (INIS)

    Ando, Yoshihira; Yoshioka, Kenichi; Mitsuhashi, Ishi; Sakurada, Koichi; Sakurai, Shungo

    2003-01-01

    Monte Carlo burnup calculation code system has been developed to evaluate accurate various quantities required in the backend field. From the Actinide Research in a Nuclear Element (ARIANE) program, by using, the measured nuclide compositions of fuel rods in the fuel assemblies irradiated in the commercial Netherlands BWR, the analyses have been performed for the code system verification. The code system developed in this paper has been verified through analysis for MOX and UO2 fuel rods. This system enables to reduce large margin assumed in the present criticality analysis for LWR spent fuels. (J.P.N.)

  3. Burn-up TRIGA Mark II benchmark experiment

    International Nuclear Information System (INIS)

    Persic, A.; Ravnik, M.; Zagar, T.

    1998-01-01

    Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

  4. Symmetry- and Solvent-Dependent Photophysics of Fluorenes Containing Donor and Acceptor Groups (Postprint)

    Science.gov (United States)

    2014-07-01

    for exciplex forma- tion.29,30 Solvent-dependent studies by Belfield et al. on similar dyes31 and BT2F32 revealed strong solvatochromism on the emission...Swiger, R. N.; Orseno, D. L.; Haley, J. E.; Cooper, T. M.; Tan, L.-S. Exciplex Formation in Solid State Blends of Charge-Transfer-Type Afx Dyes and...Tan, L.-S. Exciplex Formation in Blended Spin-Cast Films of Fluorene-Linked Dyes and Bisphthalimide Quenchers. J. Phys. Chem. A 2013, 117, 3909−3917

  5. Renormalization Group Evolution of the Standard Model Dimension Six Operators II: Yukawa Dependence

    CERN Document Server

    Jenkins, Elizabeth E; Trott, Michael

    2014-01-01

    We calculate the complete order y^2 and y^4 terms of the 59 x 59 one-loop anomalous dimension matrix for the dimension-six operators of the Standard Model effective field theory, where y is a generic Yukawa coupling. These terms, together with the terms of order lambda, lambda^2 and lambda y^2 depending on the Standard Model Higgs self-coupling lambda which were calculated in a previous work, yield the complete one-loop anomalous dimension matrix in the limit of vanishing gauge couplings. The Yukawa contributions result in non-trivial flavor mixing in the various operator sectors of the Standard Model effective theory.

  6. Finite difference solution of the time dependent neutron group diffusion equations

    International Nuclear Information System (INIS)

    Hendricks, J.S.; Henry, A.F.

    1975-08-01

    In this thesis two unrelated topics of reactor physics are examined: the prompt jump approximation and alternating direction checkerboard methods. In the prompt jump approximation it is assumed that the prompt and delayed neutrons in a nuclear reactor may be described mathematically as being instantaneously in equilibrium with each other. This approximation is applied to the spatially dependent neutron diffusion theory reactor kinetics model. Alternating direction checkerboard methods are a family of finite difference alternating direction methods which may be used to solve the multigroup, multidimension, time-dependent neutron diffusion equations. The reactor mesh grid is not swept line by line or point by point as in implicit or explicit alternating direction methods; instead, the reactor mesh grid may be thought of as a checkerboard in which all the ''red squares'' and '' black squares'' are treated successively. Two members of this family of methods, the ADC and NSADC methods, are at least as good as other alternating direction methods. It has been found that the accuracy of implicit and explicit alternating direction methods can be greatly improved by the application of an exponential transformation. This transformation is incompatible with checkerboard methods. Therefore, a new formulation of the exponential transformation has been developed which is compatible with checkerboard methods and at least as good as the former transformation for other alternating direction methods

  7. Nursing dependency, diagnosis-related groups, and length of hospital stay

    OpenAIRE

    Halloran, Edward J.; Kiley, Marylou

    1987-01-01

    Most efforts to modify the diagnosis-related group (DRG) case classification system focus on variables related to medical management. In this study, we investigated the separate but related natures of medicine and nursing by examining 1,288 adult medical and surgical patients in an urban teaching hospital. The complexity of medical treatment was measured by use of the DRG relative cost weight. The nursing indicator was derived from a set of nursing diagnoses. We found that the DRG cost weight...

  8. Preparation of computer codes for analyzing sensitivity coefficients of burnup characteristics (2) (Contract research, translated document)

    International Nuclear Information System (INIS)

    Hanaki, Hiroshi; Sanda, Toshio; Ohashi, Masahisa

    2008-10-01

    To develop nuclear design of LMFBR cores, they are important subjects of research and development to improve the accuracy in nuclear design of large LMFBR cores and to design highly efficient core more rationally. The adjusted nuclear cross-sections library has been made by being reflected the result of critical experiment of the JUPITER, etc. effectively as much as possible. And the distinct improvement of the accuracy in nuclear design of large LMFBR cores has been achieved. In the design of large LMFBR cores, however, it is important to accurately estimate not only nuclear characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. Therefore, it is thought to improve the prediction accuracy for burnup characteristics using many burnup data of 'Joyo' effectively. It is thought the best way to adjust cross sections using sensitivity coefficients of burnup characteristics to utilize burnup data of 'Joyo'. It is able to know the accuracy quantitatively for burnup characteristics of large LMFBR by analyzing the sensitivity coefficients. Therefore in this work computer codes for analyzing sensitivity coefficients of burnup characteristics had been prepared since 1992. In 1992 cross-section adjustment was done by using the data of 'Joyo' and the effect was studied. In this year the adequacy of the codes was studied with a view of applying of design of large LMFBR cores. The results are as follows: (1) The computer codes which could analyze sensitivity coefficients of burnup characteristics taking into consideration plural cycles and refueling were prepared, therefore it came of be able to adjust cross sections using burnup data and to estimate the accuracy for design of large LMFBR cores. The characteristics are not only burnup reactivity loss, breeding ratio but also number density, criticality, reactivity worth, reaction rate ratio, and reaction rate

  9. Benefits of Group Foraging Depend on Prey Type in a Small Marine Predator, the Little Penguin.

    Science.gov (United States)

    Sutton, Grace J; Hoskins, Andrew J; Arnould, John P Y

    2015-01-01

    Group foraging provides predators with advantages in over-powering prey larger than themselves or in aggregating small prey for efficient exploitation. For group-living predatory species, cooperative hunting strategies provide inclusive fitness benefits. However, for colonial-breeding predators, the benefit pay-offs of group foraging are less clear due to the potential for intra-specific competition. We used animal-borne cameras to determine the prey types, hunting strategies, and success of little penguins (Eudyptula minor), a small, colonial breeding air-breathing marine predator that has recently been shown to display extensive at-sea foraging associations with conspecifics. Regardless of prey type, little penguins had a higher probability of associating with conspecifics when hunting prey that were aggregated than when prey were solitary. In addition, success was greater when individuals hunted schooling rather than solitary prey. Surprisingly, however, success on schooling prey was similar or greater when individuals hunted on their own than when with conspecifics. These findings suggest individuals may be trading-off the energetic gains of solitary hunting for an increased probability of detecting prey within a spatially and temporally variable prey field by associating with conspecifics.

  10. Framatome-ANP extended burnup experience and views on LWR fuels

    International Nuclear Information System (INIS)

    Esteve, B.; Gueldner, R.; Hoffman, R.; Watteau, M.

    2002-01-01

    In every sense of the term, nuclear fuel forms the core of nuclear power plants (NPPs). Although there are many equipment items important for their safety function or for their participation in NPP availability, the fuel, in essence renewable, is one of the key elements which have to be acted upon if utilities are to be helped to fulfil their mission of generating power in total safety and supplying the kWh to their customers at the best price. Nuclear fuel is also the core business of the Framatome-ANP Fuel Business Group: pooling and rationalising the available skills - technical, cultural and human - supplied by each of the partners forms a challenge which it is up to each and every one to meet in a cooperative spirit. This paper gives an outline of the company's extended burnup experience, current R and D, and its plans for the future. (author)

  11. The ICRP task group respiratory tract model - an age-dependent dosimetric model for general application

    International Nuclear Information System (INIS)

    Bailey, M.R.; Birchall, A.

    1992-01-01

    The ICRP Task Group on Human Respiratory Tract Models for Radiological Protection has developed a revised dosimetric model for the respiratory tract. Papers outlining the model, and describing each aspect of it were presented at the Third International Workshop on Respiratory Tract Dosimetry (Albuquerque 1-3 July 1990), the Proceedings of which were recently published in Radiation Protection Dosimetry Volume 38 Nos 1-3 (1991). Since the model had not changed substantially since the Workshop at Albuquerque, only a summary of the paper presented at Schloss Elmau is included in these Proceedings. (author)

  12. 3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

    International Nuclear Information System (INIS)

    Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

    1974-01-01

    1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

  13. Thermal property change of MOX and UO{sub 2} irradiated up to high burnup of 74 GWd/t

    Energy Technology Data Exchange (ETDEWEB)

    Nakae, Nobuo, E-mail: nakae-nobuo@jnes.go.jp [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Kurematsu, Shigeru; Kosaka, Yuji [Nuclear Development Corporation (NDC), 622-12, Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Yoshino, Aya; Kitagawa, Takaaki [Mitsubishi Nuclear Fuel Co., LTD. (MNF), 12-1, Yurakucho 1-Chome, Chiyoda-ku, Tokyo 100-0006 (Japan)

    2013-09-15

    Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO{sub 2} fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO{sub 2}. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO{sub 2} is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO{sub 2} at high burnup under the condition that the pellet–cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO{sub 2} before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO{sub 2}. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

  14. Computer programs for TRIGA calibration, burnup evaluation, and bookkeeping

    International Nuclear Information System (INIS)

    Nelson, George W.

    1978-01-01

    Several computer programs have been developed at the University of Arizona to assist the direction and operation of the TRIGA Reactor Laboratory. The programs fall into the following three categories: 1. Programs for calculation of burnup of each fuel element in the reactor core, for maintaining an inventory of fuel element location and fissile content at any time, and for evaluation of the reactivity effects of burnup or proposed fuel element rearrangement in the core. 2. Programs for evaluation, function fitting, and tabulation of control rod measurements. 3. Bookkeeping programs to summarize and tabulate reactor runs and irradiations according to time, energy release, purpose, responsible party, etc. These summarized data are reported in an annual operating report for the facility. The use of these programs has saved innumerable hours of repetitious work, assuring more accurate, objective results, and requiring a minimum of effort to repeat calculations when input data are modified. The programs are written in FORTRAN-IV, and have been used on a CDC-6400 computer. (author)

  15. chemical determination of burnup ratio in nuclear fuels

    International Nuclear Information System (INIS)

    Guereli, L.

    1997-01-01

    Measurements of the extent of fission are important to determine the irradiation performance of a nuclear fuel. The energy released per unit mass of uranium (burnup) can be determined from measurement of the percent of heavy atoms that have fissioned during irradiation.The preferred method for this determination is choosing a suitable fission monitor (usually ''1''4''8Nd) and its determination after separation from the fuel matrix. In thermal reactor fuels where the only heavy element in the starting material is uranium, uranium depletion can be used for burnup determination. ''2''3''5U depletion method requires measurement of uranium isotopic ratios of both irradiated and unirradiated fuel. Isotopic ratios can be determined by thermal ionization mass spectrometer following separation of uranium from the fuel matrix. Separation procedures include solvent extraction, ion exchange and anion exchange chromatography. Another fission monitor used is ''1''3''9La determination by HPLC. Because La is monoisotopic (''1''3''9La) in the fuel, it can be determined by chemical analysis techniques

  16. CEXE INCEXE, 1 Group 3-D Time-Dependent Xe Oscillations in X-Y-Z Geometry with Feedback

    International Nuclear Information System (INIS)

    Kern, R.; Pacino, S.

    1973-01-01

    1 - Description of problem or function: CEXE solves the three- dimensional xyz time-dependent xenon spatial oscillation problem using a modified one energy group theory and a nodal representation. 2 - Method of solution: CEXE solves for the spatial neutron source distribution with coupled Doppler and moderator temperature feedbacks. The time dependence of the iodine and xenon concentrations are based on the assumption of constant power during each time-step interval. 3 - Restrictions on the complexity of the problem: Maximum reactor core size representation is restricted to a nodal configuration of 19 x 19 x 10 in the x, y, z directions, respectively

  17. Evolutionary Trails of Plant Group II Pyridoxal Phosphate-Dependent Decarboxylase Genes.

    Science.gov (United States)

    Kumar, Rahul

    2016-01-01

    Type II pyridoxal phosphate-dependent decarboxylase (PLP_deC) enzymes play important metabolic roles during nitrogen metabolism. Recent evolutionary profiling of these genes revealed a sharp expansion of histidine decarboxylase genes in the members of Solanaceae family. In spite of the high sequence homology shared by PLP_deC orthologs, these enzymes display remarkable differences in their substrate specificities. Currently, limited information is available on the gene repertoires and substrate specificities of PLP_deCs which renders their precise annotation challenging and offers technical challenges in the immediate identification and biochemical characterization of their full gene complements in plants. Herein, we explored their evolutionary trails in a comprehensive manner by taking advantage of high-throughput data accessibility and computational approaches. We discussed the premise that has enabled an improved reconstruction of their evolutionary lineage and evaluated the factors offering constraints in their rapid functional characterization, till date. We envisage that the synthesized information herein would act as a catalyst for the rapid exploration of their biochemical specificity and physiological roles in more plant species.

  18. Determination of the burn-up of TRIGA fuel elements by calculation with new TRIGLAV program

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.

    1996-01-01

    The results of fuel element burn-up calculations with new TRIGLAV program are presented. TRIGLAV program uses two dimensional model. Results of calculation are compared to results calculated with program, which uses one dimensional model. The results of fuel element burn-up measurements with reactivity method are presented and compared with the calculated results. (author)

  19. Application of reactivity method to MTR fuel burn-up measurement

    International Nuclear Information System (INIS)

    Zuniga, A.; Ravnik, M.; Cuya, R.

    2001-01-01

    Fuel element burn-up has been measured for the first time by reactivity method in a MTR reactor. The measurement was performed in RP-10 reactor of Peruvian Institute for Nuclear Energy (IPEN) in Lima. It is a pool type 10MW material testing reactor using standard 20% enriched uranium plate type fuel elements. A fresh element and an element with well defined burn-up were selected as reference elements. Several elements in the core were selected for burn-up measurement. Each of them was replaced in its original position by both reference elements. Change in excess reactivity was measured using control rod calibration curve. The burn-up reactivity worth of fuel elements was plotted as a function of their calculated burnup. Corrected burn-up values of the measured fuel elements were calculated using the fitting function at experimental reactivity for all elements. Good agreement between measured and calculated burn-up values was observed indicating that the reactivity method can be successfully applied also to MTR fuel element burn-up determination.(author)

  20. Determination of axial profit performed burnup credit by SCALE 4.3-system

    International Nuclear Information System (INIS)

    Miro, R.; Verdu, G.; Munoz-Cobo, J. L.

    1998-01-01

    SCALE 4.3 is a modular code system designed for realizing standard computational analysis for licensing evaluation. Since now, spent fuel storage pools criticality analysis have been done considering this fuel as fresh, with its maximum enrichment. With burnup credit we can obtain cheaper and compact configurations. The procedure for calculating a spent fuel storage consists of a burnup calculation plus a criticality calculation. We can perform a conservative approximation for the burnup calculations using 1-D results, but, besides the geometry configurations for the 3-D criticality calculation. we need an appropriate approximation to model the burnup axial variation. We assume that for a burnup profile set, the most conservative profile is between the lower and the upper range of this profile, set. We consider only combinations of the maximum and minimum burnup in each axial region, for each burnup range. This gives an estimation of the different burnup shapes effect and the general characteristics of the most conservative shape. (Author) 6 refs

  1. Evaluation of Isotopic Measurements and Burn-up Value of Sample GU3 of ARIANE Project

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Rodriguez Rivada, A.

    2014-07-01

    Estimation of the burn-up value of irradiated fuel and its isotopic composition are important for criticality analysis, spent fuel management and source term estimation. The practical way to estimate the irradiated fuel composition and burn.up value is calculation with validated code and nuclear data. Such validation of the neutronic codes and nuclear data requires the benchmarking with measured values. (Author)

  2. Determination of nuclear fuel burn-up using mass spectrometric techniques

    International Nuclear Information System (INIS)

    Saha, B.; Bagyalakshmi, R.; Periaswami, G.; Kavimandan, V.D.; Chitambar, S.A.; Jain, H.C.; Mathews, C.K.

    1977-01-01

    Determination of burn-up using a stable fission product monitor such as 148 Nd and heavy elements, determined by isotope dilution mass spectrometry gives the most accurate data. This report describes the work carried out to standardise the conditions for burn-up determination. Some typical results are given. (author)

  3. The application of burnup credit for spent fuel operations in the United Kingdom

    International Nuclear Information System (INIS)

    Bowden, R.

    1998-01-01

    This paper begins by outlining the structure of the nuclear industry in the United Kingdom. It then sets out the methodology of burnup credit, and provides a brief discussion of the validation and robustness of the calculational route. This leads to a description of both the current and intended applications of burnup credit in the United Kingdom. (author)

  4. Simulation of High Burnup Structure in UO2 Using Potts Model

    International Nuclear Information System (INIS)

    Oh, Jae Yong; Koo, Yang Hyun; Lee, Byung Ho

    2009-01-01

    The evolution of a high burnup structure (HBS) in a light water reactor (LWR) UO 2 fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the UO 2 matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels

  5. The relevance of axial burn-up profiles for the criticality safety analysis of spent nuclear fuel in a final repository

    International Nuclear Information System (INIS)

    Kilger, R.; Gmal, B.; Moser, E.F.

    2008-01-01

    Due to inhomogeneous neutron flux and moderator density distributions in the reactor core, the burn-up of a nuclear fuel assembly is not homogeneous but shows an axial distribution, typically with lower partial burn-up and thus higher remaining reactivity at the fuel ends in particular at the assembly top end. Beyond a burn-up of about 15 to 20 GWd/tHM, the multiplication factor K of the whole assembly is dominated by this lower-burnt end regions, and is usually higher than for assuming a homogeneous uniform distribution of the averaged burn-up. This behaviour commonly referred to as positive ''end effect'' is well known in burn-up credit considerations for transportation and storage casks and is being investigated also in the context of criticality analyses for final disposition of spent nuclear fuel. Sign and value of the end effect depend on several parameters. Based on a generic model one may not conclude that criticality in a final repository is a likely or expected event, but nevertheless it draws the attention to the fact that criticality is not excluded per se but has to be considered in the analysis and probably has to be encountered by certain appropriate measures, maybe e.g. by limitation of the amount of fissile material inside one single cask, or a rigorous prove for prevention of water ingress. The authors also conclude that the higher partial reactivity of the fuel ends has to be accounted for carefully in more realistic analyses of post-closure scenarios with respect to criticality safety.

  6. Chemical modification and pH dependence of kinetic parameters to identify functional groups in a glucosyltransferase from Strep. Mutans

    International Nuclear Information System (INIS)

    Bell, J.E.; Leone, A.; Bell, E.T.

    1986-01-01

    A glucosyltransferase, forming a predominantly al-6 linked glucan, was partially purified from the culture filtrate of S. mutans GS-5. The kinetic properties of the enzyme, assessed using the transfer of 14 C glucose from sucrose into total glucan, were studied at pH values from pH 3.5 to 6.5. From the dependence of km on pH, a group with pKa = 5.5 must be protonated to maximize substrate binding. From plots of V/sub max/ vs pH two groups, with pKa's of 4.5 and 5.5 were indicated. The results suggest the involvement of either two carboxyl groups (one protonated, one unprotonated in the native enzyme) or a carboxyl group (unprotonated) and some other protonated group such as histidine, cysteine. Chemical modification studies showed that Diethylyrocarbonate (histidine specific) had no effect on enzyme activity while modification with p-phydroxy-mercuribenzoate or iodoacetic acid (sulfhydryl reactive) and carbodimide reagents (carboxyl specific) resulted in almost complete inactivation. Activity loss was dependent upon time of incubation and reagent concentration. The disaccharide lylose, (shown to be an inhibitor of the enzyme with similar affinity to sucrose) offers no protection against modification by the sulfhydryl reactive reagents

  7. Preparation of data relevant to ''Equivalent Uniform Burnup'' and Equivalent Initial Enrichment'' for burnup credit evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan)

    2001-11-01

    Based on the PWR spent fuel composition data measured at JAERI, two kinds of simplified methods such as ''Equivalent Uniform Burnup'' and ''Equivalent Initial Enrichment'' have been introduced. And relevant evaluation curves have been prepared for criticality safety evaluation of spent fuel storage pool and transport casks, taking burnup of spent fuel into consideration. These simplified methods can be used to obtain an effective neutron multiplication factor for a spent fuel storage/transportation system by using the ORIGEN2.1 burnup code and the KENO-Va criticality code without considering axial burnup profile in spent fuel and other various factors introducing calculated errors. ''Equivalent Uniform Burnup'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis, in which the experimentally obtained isotopic composition together with a typical axial burnup profile and various factors such as irradiation history are considered on the conservative side. On the other hand, Equivalent Initial Enrichment'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis such as above when it is used in the so called fresh fuel assumption. (author)

  8. The effectiveness of group dietary counselling among non insulin dependent diabetes mellitus (NIDDM) patients in resettlement scheme areas in Malaysia.

    Science.gov (United States)

    Talib, R; Ali, O; Arshad, F; Kadir, K A

    1997-06-01

    A study was undertaken in FELDA (Federal Land Development Authority) resettlement scheme areas in Pahang, Malaysia, to determine the effectiveness of group dietary counselling in motivating diabetic patients to achieve good dietary habits, and weight and diabetes control. Sixty-one non-insulin dependent diabetes mellitus (NIDDM) patients were randomly assigned to either the experimental or control group. The experimental group received six sessions of group dietary counselling over 5 months and the control group received mass media diabetes-educational program during the same period. The one hour group dietary counselling sessions discussed general knowledge of diabetes, food groups for meal planning, the importance of dietary fibre-rich foods, types of fat in food, exercise and weight control. The experimental group met monthly with a dietitian as a counsellor. Effectiveness was assessed by improvement in food choice, and decline in percentage glycated haemoglobin (total HbA1) or body mass index (BMI). Measurements were made at a baseline visit, every two months during the six month program, and six months afterwards. Patients in the experimental group improved their food choices, resulting in a healthier diet high in unrefined carbohydrates and dietary fibre rich foods, and low in fat. There were significant reductions of their percentage total HbA1 levels and BMI following the counselling sessions, which decreased further six months after the program compared with patients in the control group. Thus group dietary counselling is effective in motivating NIDDM patients to achieve better food choice, and related weight and glycaemic control in a Malaysian setting.

  9. Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Sakr, A.M.; Amin, E.H.

    2011-01-01

    Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

  10. A survey of previous and current industry-wide efforts regarding burnup credit

    International Nuclear Information System (INIS)

    Jones, R.H.

    1989-01-01

    Sandia has examined the matter of burnup credit from the perspective of physics, logistics, risk, and economics. A limited survey of the nuclear industry has been conducted to get a feeling for the actual application of burnup credit. Based on this survey, it can be concluded that the suppliers of spent fuel storage and transport casks are in general agreement that burnup credit offers the potential for improvements in cask efficiency without increasing the risk of accidental criticality. The actual improvement is design-specific but limited applications have demonstrated that capacity increases in the neighborhood of 20 percent are not unrealistic. A number of these vendors acknowledge that burnup credit has not been reduced to practice in cask applications and suggest that operational considerations may be more important to regulatory acceptance than to the physics. Nevertheless, the importance of burnup credit to the nuclear industry as a cask design and analysis tool has been confirmed by this survey

  11. Regulatory status of burnup credit for storage and transport of spent fuel in Germany

    International Nuclear Information System (INIS)

    Neuber, J.C.; Schweer, H.H.; Johann, H.G.

    2001-01-01

    This paper describes the regulatory status of burnup credit applications to pond storage and dry-cask transport and storage of spent fuel in Germany. Burnup credit for wet storage of LWR fuel at nuclear power plants has to comply with the newly developed safety standard DIN 25471. This standard establishes the safety requirements for burnup credit criticality safety analysis of LWR fuel storage ponds and gives guidance on meeting these requirements. Licensing evaluations of dry transport systems are based on the application of the IAEA Safety Standards Series No.ST-1. However, because of the fact that burnup credit for dry-cask transport becomes more and more inevitable due to increasing initial enrichment of the fuel, and because of the increasing importance of dry-cask storage in Germany, the necessity of giving regulatory guidance on applying burnup credit to dry-cask transport and storage is seen. (author)

  12. Assessment of US NRC fuel rod behavior codes to extended burnup

    International Nuclear Information System (INIS)

    Laats, E.T.; Croucher, D.W.; Haggag, F.M.

    1982-01-01

    The purpose of this paper is to report the status of assessing the capabilities of the NRC fuel rod performance codes for calculating extended burnup rod behavior. As part of this effort, a large spectrum of fuel rod behavior phenomena was examined, and the phenomena deemed as being influential during extended burnup operation were identified. Then, the experiment data base addressing these identified phenomena was examined for availability and completeness at extended burnups. Calculational capabilities of the NRC's steady state FRAPCON-2 and transient FRAP-T6 fuel rod behavior codes were examined for each of the identified phenomenon. Parameters calculated by the codes were compared with the available data base, and judgments were made regarding model performance. Overall, the FRAPCON-2 code was found to be moderately well assessed to extended burnups, but the FRAP-T6 code cannot be adequately assessed until more transient high burnup data are available

  13. Cell verification of parallel burnup calculation program MCBMPI based on MPI

    International Nuclear Information System (INIS)

    Yang Wankui; Liu Yaoguang; Ma Jimin; Wang Guanbo; Yang Xin; She Ding

    2014-01-01

    The parallel burnup calculation program MCBMPI was developed. The program was modularized. The parallel MCNP5 program MCNP5MPI was employed as neutron transport calculation module. And a composite of three solution methods was used to solve burnup equation, i.e. matrix exponential technique, TTA analytical solution, and Gauss Seidel iteration. MPI parallel zone decomposition strategy was concluded in the program. The program system only consists of MCNP5MPI and burnup subroutine. The latter achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, and data exchanging with MCNP5MPI. Also, the program was verified with the pressurized water reactor (PWR) cell burnup benchmark. The results show that it,s capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

  14. Application of burnup credit in spent fuel management at Russian NPPs

    International Nuclear Information System (INIS)

    Koulikov, V.I.; Makarchuk, T.F.; Tikhonov, N.S.

    1998-01-01

    The article concerns implementation of burnup credit in spent fuel storage and transportation. Some of the problems with increased enrichment fuel can be resolved by use of modified transport methodology. Such as shipping in gas-filled casks only, reduced number of assemblies in casks, etc. However, the use of modified schemes of transportation results in essential financial losses. An actinide-only burnup credit is taken into account in most part of criticality calculations, and a parameter limiting loading of spent fuel in the cask or the repository is the avenge value of burnup on an assembly. The main method of burnup depth definition is its defect measurement. A short description of devices for measurement as well as some technical results of suing burnup credit approach in storage and transport are given. (author)

  15. HIC1 links retinoic acid signalling to group 3 innate lymphoid cell-dependent regulation of intestinal immunity and homeostasis

    Science.gov (United States)

    Antignano, Frann; Korinek, Vladimir; Underhill, T. Michael

    2018-01-01

    The intestinal immune system must be able to respond to a wide variety of infectious organisms while maintaining tolerance to non-pathogenic microbes and food antigens. The Vitamin A metabolite all-trans-retinoic acid (atRA) has been implicated in the regulation of this balance, partially by regulating innate lymphoid cell (ILC) responses in the intestine. However, the molecular mechanisms of atRA-dependent intestinal immunity and homeostasis remain elusive. Here we define a role for the transcriptional repressor Hypermethylated in cancer 1 (HIC1, ZBTB29) in the regulation of ILC responses in the intestine. Intestinal ILCs express HIC1 in a vitamin A-dependent manner. In the absence of HIC1, group 3 ILCs (ILC3s) that produce IL-22 are lost, resulting in increased susceptibility to infection with the bacterial pathogen Citrobacter rodentium. Thus, atRA-dependent expression of HIC1 in ILC3s regulates intestinal homeostasis and protective immunity. PMID:29470558

  16. Interaction of PAMAM dendrimers with bovine insulin depends on nanoparticle end-groups

    International Nuclear Information System (INIS)

    Nowacka, Olga; Milowska, Katarzyna; Bryszewska, Maria

    2015-01-01

    We have looked at the interactions between polyamidoamine (PAMAM) dendrimers with different terminal groups (−COOH, −NH 2 , −OH) and bovine insulin. The influence of PAMAM dendrimers on insulin was tested by measuring zeta potential and fluorescence quenching. The secondary structure of insulin in the presence of dendrimers was examined by circular dichroism. The effect of dendrimers on dithiotreitol-induced aggregation of insulin was investigated by spectrophotometry. Dendrimers quenched the fluorescence of insulin, but did not change its secondary structure. Thus dendrimers neither induce hormone aggregation nor inhibit the aggregation process induced by dithiotreitol (DTT), except at 0.01 µmol/l. Dendrimers–insulin interactions are mainly electrostatic. - Highlight: • The interactions between PAMAM dendrimers and insulin were investigated. • The PAMAM dendrimers can quench the fluorescence of insulin. • The PAMAM dendrimers did not change the secondary structure of insulin. • Dendrimers did not induce aggregation of hormone. • Dendrimers–insulin interaction is mainly electrostatic

  17. Dependence of micelle size and shape on detergent alkyl chain length and head group.

    Directory of Open Access Journals (Sweden)

    Ryan C Oliver

    Full Text Available Micelle-forming detergents provide an amphipathic environment that can mimic lipid bilayers and are important tools for solubilizing membrane proteins for functional and structural investigations in vitro. However, the formation of a soluble protein-detergent complex (PDC currently relies on empirical screening of detergents, and a stable and functional PDC is often not obtained. To provide a foundation for systematic comparisons between the properties of the detergent micelle and the resulting PDC, a comprehensive set of detergents commonly used for membrane protein studies are systematically investigated. Using small-angle X-ray scattering (SAXS, micelle shapes and sizes are determined for phosphocholines with 10, 12, and 14 alkyl carbons, glucosides with 8, 9, and 10 alkyl carbons, maltosides with 8, 10, and 12 alkyl carbons, and lysophosphatidyl glycerols with 14 and 16 alkyl carbons. The SAXS profiles are well described by two-component ellipsoid models, with an electron rich outer shell corresponding to the detergent head groups and a less electron dense hydrophobic core composed of the alkyl chains. The minor axis of the elliptical micelle core from these models is constrained by the length of the alkyl chain, and increases by 1.2-1.5 Å per carbon addition to the alkyl chain. The major elliptical axis also increases with chain length; however, the ellipticity remains approximately constant for each detergent series. In addition, the aggregation number of these detergents increases by ∼16 monomers per micelle for each alkyl carbon added. The data provide a comprehensive view of the determinants of micelle shape and size and provide a baseline for correlating micelle properties with protein-detergent interactions.

  18. Modification of UO2 grain re-crystallization temperature in function of burn-up as a base for Vitanza experimental curve reconstruction

    International Nuclear Information System (INIS)

    Szuta, M.; Dąbrowski, L.

    2013-01-01

    Crossing the experimental critical fuel temperature dependent on burn-up, an onset of fission gas burst release is observed. This observed phenomena can be explained by assumption that the fission gas immobilization in the uranium dioxide irradiated to a fluency of greater than 10 19 fissions/cm 3 is mainly due to radiation induced chemical activity. Application of the “ab initio” method show that the bond energy of Xenon and Krypton is equal to –1.23 eV, and –3.42 eV respectively. Assuming further that the gas chemically bound can be released mainly in the process of re-crystallization and modifying the differential equation of Ainscough of grain growth by including the burn-up dependence and the experimental data of limiting grain size in function of the fuel temperature for the un-irradiated and irradiated fuel we can re-construct the experimental curve of Vitanza. (authors)

  19. Measurement of gamma attenuation coefficients in UO2 and zirconium for self-absorption corrections of burn-up determination

    International Nuclear Information System (INIS)

    Podest, M.; Klima, J.; Stecher, P.; Stecherova, E.

    1978-01-01

    UO 2 pellets from ALUOX fuel elements were used in measuring the absorption coefficient of gamma radiation in UO 2 . The results of measurements of the energy dependence of the linear absorption coefficient (within 622 to 796 keV) and of the dependence on pellet density showed that in the given density interval the absorption coefficient was almost constant. The density interval was chosen to be typical for pellet fuel used in water cooled and water moderated power reactors. The results are also shown of the dependence of the mass absorption coefficient of gamma radiation in Zr on radiation energy and compared with the mass absorption coefficient of Mo; these also showed the independence of the absorption coefficient on density. The linear and mass absorption coefficients of UO 2 are considerably high and correspond approximately to the absorption coefficient of lead. For the measured energy range the variation of absorption coefficient is about 40%, which causes errors in burnup determination. The efficiency was also determined of Ge(Li) detectors for the energy range 0.5 to 1.2 MeV. The determination of the above coefficients was used for improving the gamma fuel scanning technique in determining the activity and burnup of spent fuel elements. (J.P.)

  20. Determination of burn-up of irradiated nuclear fuels using mass spectrometry

    International Nuclear Information System (INIS)

    Jagadish Kumar, S.; Telmore, V.M.; Shah, R.V.; Sasi Bhushan, K.; Paul, Sumana; Kumar, Pranaw; Rao, Radhika M.; Jaison, P.G.

    2017-01-01

    Burn-up defined as the atom percent fission, is a vital parameter used for assessing the performance of nuclear fuel during its irradiation in the reactor. Accurate data on the actinide isotopes are also essential for the reliable accountability of nuclear materials and for nuclear safeguards. Both destructive and non-destructive methods are employed in the post-irradiation analysis for the burn-up measurements. Though non-destructive methods are preferred from the point view of remote handling of irradiated fuels with high radioactivity, they do not provide the high accuracy as achieved by the chemical analysis methods. Thus destructive radiochemical and chemical analyses are still the established reference methods for accurate and reliable burn-up determination of irradiated nuclear fuels. In the destructive method, burn-up of irradiated nuclear fuel is determined by correlating the amount of a fission product formed during irradiation with that of heavy elements. Thus the destructive experimental determination of burn-up involves the dissolution of irradiated fuel samples followed by the separation and determination of heavy elements and fission product(s) to be used as burn-up monitor(s). Another approach for the experimental determination of burn-up is based on the changes in the abundances of the heavy element isotopes. A widely accepted method for burn-up determination is based on stable "1"4"8Nd and "1"3"9La as burn-up monitors. Several properties such as non-volatility, nearly same yields for thermal fissions of "2"3"5U and "2"3"9Pu etc justifies the selection of "1"4"8Nd as a burn-up monitor

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

    International Nuclear Information System (INIS)

    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

  2. Simulation of triton burn-up in JET plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Loughlin, M J; Balet, B; Jarvis, O N; Stubberfield, P M [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    This paper presents the first triton burn-up calculations for JET plasmas using the transport code TRANSP. Four hot ion H-mode deuterium plasmas are studied. For these discharges, the 2.5 MeV emission rises rapidly and then collapses abruptly. This phenomenon is not fully understood but in each case the collapse phase is associated with a large impurity influx known as the ``carbon bloom``. The peak 14 MeV emission occurs at this time, somewhat later than that of the 2.5 MeV neutron peak. The present results give a clear indication that there are no significant departures from classical slowing down and spatial diffusion for tritons in JET plasmas. (authors). 7 refs., 3 figs., 1 tab.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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).

  4. User's manual for the reactor burnup system, REBUS

    International Nuclear Information System (INIS)

    Olson, A.P.; Regis, J.P.; Meneley, D.A.; Hoover, L.J.

    1972-01-01

    A user's manual for the REBUS System (REactor BUrnup System) is presented. Its primary purpose is to provide sufficient information about the REBUS capability to the user to ensure its efficient utilization. The current REBUS System either solves for the infinite time (equilibrium) operating conditions of a recycle system under fixed conditions, or solves for operating conditions during a single time step (non-equilibrium). The capability of studying various in-reactor fuel management and ex-reactor fuel management schemes has been included. REBUS has been operated with one- and two-dimensional diffusion theory neutronics solutions up to the present time. The model was specifically designed for extension to other neutronics models such as three-dimensional diffusion or transport theory and direct or synthesis solutions

  5. Manufacturing Data Uncertainties Propagation Method in Burn-Up Problems

    Directory of Open Access Journals (Sweden)

    Thomas Frosio

    2017-01-01

    Full Text Available A nuclear data-based uncertainty propagation methodology is extended to enable propagation of manufacturing/technological data (TD uncertainties in a burn-up calculation problem, taking into account correlation terms between Boltzmann and Bateman terms. The methodology is applied to reactivity and power distributions in a Material Testing Reactor benchmark. Due to the inherent statistical behavior of manufacturing tolerances, Monte Carlo sampling method is used for determining output perturbations on integral quantities. A global sensitivity analysis (GSA is performed for each manufacturing parameter and allows identifying and ranking the influential parameters whose tolerances need to be better controlled. We show that the overall impact of some TD uncertainties, such as uranium enrichment, or fuel plate thickness, on the reactivity is negligible because the different core areas induce compensating effects on the global quantity. However, local quantities, such as power distributions, are strongly impacted by TD uncertainty propagations. For isotopic concentrations, no clear trends appear on the results.

  6. Validating analysis methodologies used in burnup credit criticality calculations

    International Nuclear Information System (INIS)

    Brady, M.C.; Napolitano, D.G.

    1992-01-01

    The concept of allowing reactivity credit for the depleted (or burned) state of pressurized water reactor fuel in the licensing of spent fuel facilities introduces a new challenge to members of the nuclear criticality community. The primary difference in this analysis approach is the technical ability to calculate spent fuel compositions (or inventories) and to predict their effect on the system multiplication factor. Isotopic prediction codes are used routinely for in-core physics calculations and the prediction of radiation source terms for both thermal and shielding analyses, but represent an innovation for criticality specialists. This paper discusses two methodologies currently being developed to specifically evaluate isotopic composition and reactivity for the burnup credit concept. A comprehensive approach to benchmarking and validating the methods is also presented. This approach involves the analysis of commercial reactor critical data, fuel storage critical experiments, chemical assay isotopic data, and numerical benchmark calculations

  7. Oxide thickness measurement for monitoring fuel performance at high burnup

    International Nuclear Information System (INIS)

    Jaeger, M.A.; Van Swam, L.F.P.; Brueck-Neufeld, K.

    1991-01-01

    For on-site monitoring of the fuel performance at high burnup, Advanced Nuclear Fuels uses the linear scan eddy current method to determine the oxide thickness of irradiated Zircaloy fuel cans. Direct digital data acquisition methods are employed to collect the data on magnetic storage media. This field-proven methodology allows oxide thickness measurements and rapid interpretation of the data during the reactor outages and makes it possible to immediately reinsert the assemblies for the next operating cycle. The accuracy of the poolside measurements and data acquisition/interpretation techniques have been verified through hot cell metallographic measurements of rods previously measured in the fuel pool. The accumulated data provide a valuable database against which oxide growth models have been benchmarked and allow for effective monitoring of fuel performance. (orig.) [de

  8. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    Chernushev, V.; Sokolov, F.

    2002-01-01

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

  9. Approach to lithium burn-up effect in lithium ceramics

    International Nuclear Information System (INIS)

    Rasneur, B.

    1994-01-01

    The lithium burn-up in Li 2 ZrO 3 is simulated by removing lithium under Li 2 O form and trapping it in high specific surface area powder while heating during 15 days or 1 month at moderate temperature so that lithium mobility be large enough without causing any sintering neither of the specimens nor of the powder. In a first treatment at 775 deg C during 1 month. 30% of the lithium content could be removed inducing a lithium concentration gradient in the specimen and the formation of a lithium-free monoclinic ZrO 2 skin. Improvements led to similar results at 650 deg C and 600 deg C, the latter temperatures are closer to the operating temperature of the ceramic breeder blanket of a fusion reactor. (author) 4 refs.; 4 figs.; 1 tab

  10. Effects of aromaticity in cations and their functional groups on the temperature dependence of low-frequency spectrum

    Science.gov (United States)

    Kakinuma, Shohei; Ramati, Sharon; Wishart, James F.; Shirota, Hideaki

    2018-05-01

    In this study, we investigate the temperature dependence of low-frequency spectra in the frequency range of 0.3-200 cm-1 for ionic liquids (ILs) whose cations possess two systematically different cyclic groups, using femtosecond Raman-induced Kerr effect spectroscopy. The target ILs are bis(trifluoromethylsulfonyl)amide [NTf2]- salts of 1-cyclohexylmethyl-1-methylpyrrolidinium [CHxmMPyrr]+, 1-cyclohexylmethyl-3-methylimidazolium [CHxmMIm]+, N-cyclohexylmethylpyridinium [CHxmPy]+, 1-benzyl-1-methylpyrrolidinium [BzMPyrr]+, 1-benzyl-3-methylimidazolium [BzMIm]+, and N-benzylpyridinium [BzPy]+ cations. The aim of this study is to better understand the effects of aromaticity in the cations' constituent groups on the temperature-dependent low-frequency spectral features of the ILs. The low-frequency spectra of these ILs are temperature dependent, but the temperature-dependent spectrum of [CHxmMPyrr][NTf2] is different from that of other ILs. While [CHxmMPyrr][NTf2] shows spectral changes with temperature in the low-frequency region below 50 cm-1, the other ILs also show spectral changes in the high-frequency region above 80 cm-1 (above 50 cm-1 in the case of [BzMPyrr][NTf2]). We conclude that the spectral change in the low-frequency region is due to both the cation and anion, while the change in the high-frequency region is attributed to the red shift of the aromatic ring librations. On the basis of the plots of the first moment of the spectra vs. temperature, we found that the first moment of the low-frequency spectrum of the IL whose cation does not have an aromatic ring is less temperature dependent than that of the other ILs. However, the intrinsic first moment, the first moment at 0 K, of the low-frequency spectrum is governed by the absence or presence of a charged aromatic group, while a neutral aromatic group does not have much influence on determining the intrinsic first moment.

  11. A Hybrid Metaheuristic Approach for Minimizing the Total Flow Time in A Flow Shop Sequence Dependent Group Scheduling Problem

    Directory of Open Access Journals (Sweden)

    Antonio Costa

    2014-07-01

    Full Text Available Production processes in Cellular Manufacturing Systems (CMS often involve groups of parts sharing the same technological requirements in terms of tooling and setup. The issue of scheduling such parts through a flow-shop production layout is known as the Flow-Shop Group Scheduling (FSGS problem or, whether setup times are sequence-dependent, the Flow-Shop Sequence-Dependent Group Scheduling (FSDGS problem. This paper addresses the FSDGS issue, proposing a hybrid metaheuristic procedure integrating features from Genetic Algorithms (GAs and Biased Random Sampling (BRS search techniques with the aim of minimizing the total flow time, i.e., the sum of completion times of all jobs. A well-known benchmark of test cases, entailing problems with two, three, and six machines, is employed for both tuning the relevant parameters of the developed procedure and assessing its performances against two metaheuristic algorithms recently presented by literature. The obtained results and a properly arranged ANOVA analysis highlight the superiority of the proposed approach in tackling the scheduling problem under investigation.

  12. Restructuring of burnup sensitivity analysis code system by using an object-oriented design approach

    International Nuclear Information System (INIS)

    Kenji, Yokoyama; Makoto, Ishikawa; Masahiro, Tatsumi; Hideaki, Hyoudou

    2005-01-01

    A new burnup sensitivity analysis code system was developed with help from the object-oriented technique and written in Python language. It was confirmed that they are powerful to support complex numerical calculation procedure such as reactor burnup sensitivity analysis. The new burnup sensitivity analysis code system PSAGEP was restructured from a complicated old code system and reborn as a user-friendly code system which can calculate the sensitivity coefficients of the nuclear characteristics considering multicycle burnup effect based on the generalized perturbation theory (GPT). A new encapsulation framework for conventional codes written in Fortran was developed. This framework supported to restructure the software architecture of the old code system by hiding implementation details and allowed users of the new code system to easily calculate the burnup sensitivity coefficients. The framework can be applied to the other development projects since it is carefully designed to be independent from PSAGEP. Numerical results of the burnup sensitivity coefficient of a typical fast breeder reactor were given with components based on GPT and the multicycle burnup effects on the sensitivity coefficient were discussed. (authors)

  13. Full Core Burn-up Calculation at JRR-3 with MVP-BURN

    International Nuclear Information System (INIS)

    Komeda, Masao; Yamamoto, Kazuyoshi; Kusunoki, Tsuyoshi

    2008-01-01

    Research reactors use a burnable poison to suppress an excess reactivity in the beginning of reactor lifetime. The JRR-3 (Japan Research Reactor No.3) has used cadmium wires of radius 0.02 cm as a burnable poison. This report describes burn-up calculations of plate fuel models and full core models with MVP-BURN, which is a burn-up calculation code using Monte Carlo method and has been developed in JAEA (Japan Atomic Energy Agency). As the results of calculations of plate models, between a model composed of one burn-up region along the radius direction and a model composed of a few burn-up regions along the radius direction, the effective absorption cross section of 113 Cd has had different tendency on reaching approximate 40. day (10000 MWd/t). And as results of calculations of full core model, it has been indicated that k eff is almost same till approximate 80. day (22000 MWd/t) between a model composed of one burn-up region along the vertical direction and a model composed of a few burn-up regions along the vertical direction. However difference of 113 Cd burn-up becomes pronounced and each k eff makes a difference after 80. day. (authors)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

    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

  15. Analyzing the BWR rod drop accident in high-burnup cores

    International Nuclear Information System (INIS)

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

    1995-01-01

    This study was undertaken for the US Nuclear Regulatory Commission to determine the fuel enthalpy during a rod drop accident (RDA) for cores with high burnup fuel. The calculations were done with the RAMONA-4B code which models the core with 3-dimensional neutron kinetics and multiple parallel coolant channels. The calculations were done with a model for a BWR/4 with fuel bundles having burnups up to 30 GWd/t and also with a model with bundle burnups to 60 GWd/t. This paper also discusses potential sources of uncertainty in calculations with high burnup fuel. One source is the ''rim'' effect which is the extra large peaking of the power distribution at the surface of the pellet. This increases the uncertainty in reactor physics and heat conduction models that assume that the energy deposition has a less peaked spatial distribution. Two other sources of uncertainty are the result of the delayed neutron fraction decreasing with burnup and the positive moderator temperature feedback increasing with burnup. Since these effects tend to increase the severity of the event, an RDA calculation for high burnup fuel will underpredict the fuel enthalpy if the effects are not properly taken into account. Other sources of uncertainty that are important come from the initial conditions chosen for the RDA. This includes the initial control rod pattern as well as the initial thermal-hydraulic conditions

  16. Determination of enrichment of recycle uranium fuels for different burnup values

    International Nuclear Information System (INIS)

    Zabunoglu, Okan H.

    2008-01-01

    Uranium (U) recovered from spent LWR fuels by reprocessing, which contains small amounts of U-236, is to be enriched before being re-irradiated as the recycle U. During the enrichment of recovered U in U-235, the mass fraction of U-236 also increases. Since the existence of U-236 in the recycle U has a negative effect on neutron economy, a greater enrichment of U-235 in the recycle U is required for reaching the same burnup as can be reached by the fresh U fuel. Two burnup values play the most important role in determining the enrichment of recycle U: (1) discharge burnup of spent fuel from which the recycle U is obtained and (2) desired discharge burnup of the recycle U fuel. A step-by-step procedure for calculating the enrichment of the recycle U as a function of these two burnup values is introduced. The computer codes MONTEBURNS and ORIGEN-S are made use of and a three-component (U-235, U-236, U-238) enrichment scheme is applied for calculating the amount of U-236 in producing the recycle U from the recovered U. As was aimed, the resulting expression is simple enough for quick/hand calculations of the enrichment of the recycle U for any given discharge burnup of spent fuel and for any desired discharge burnup of the recycle U fuel, most accurately within the range of 33,000-50,000 MWd/tonU

  17. Development and Applications of a Prototypic SCALE Control Module for Automated Burnup Credit Analysis

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2001-01-01

    Consideration of the depletion phenomena and isotopic uncertainties in burnup-credit criticality analysis places an increasing reliance on computational tools and significantly increases the overall complexity of the calculations. An automated analysis and data management capability is essential for practical implementation of large-scale burnup credit analyses that can be performed in a reasonable amount of time. STARBUCS is a new prototypic analysis sequence being developed for the SCALE code system to perform automated criticality calculations of spent fuel systems employing burnup credit. STARBUCS is designed to help analyze the dominant burnup credit phenomena including spatial burnup gradients and isotopic uncertainties. A search capability also allows STARBUCS to iterate to determine the spent fuel parameters (e.g., enrichment and burnup combinations) that result in a desired k eff for a storage configuration. Although STARBUCS was developed to address the analysis needs for spent fuel transport and storage systems, it provides sufficient flexibility to allow virtually any configuration of spent fuel to be analyzed, such as storage pools and reprocessing operations. STARBUCS has been used extensively at Oak Ridge National Laboratory (ORNL) to study burnup credit phenomena in support of the NRC Research program

  18. Evolutionary Profiling of Group II Pyridoxal-Phosphate-Dependent Decarboxylases Suggests Expansion and Functional Diversification of Histidine Decarboxylases in Tomato

    Directory of Open Access Journals (Sweden)

    Rahul Kumar

    2016-03-01

    Full Text Available Pyridoxal phosphate (PLP-dependent enzymes are one of the most important enzymes involved in plant N metabolism. Here, we explored the evolution of group II PLP-dependent decarboxylases (PLP_deC, including aromatic L-amino acid decarboxylase, glutamate decarboxylase, and histidine decarboxylase in the plant lineage. Gene identification analysis revealed a higher number of genes encoding PLP_deC in higher plants than in lower plants. Expression profiling of PLP_deC orthologs and syntelogs in (L. Heynh., pepper ( L., and tomato ( L. pointed toward conserved as well as distinct roles in developmental processes such as fruit maturation and ripening and abiotic stress responses. We further characterized a putative promoter of tomato ripening-associated gene ( operating in a complex regulatory circuit. Our analysis provides a firm basis for further in-depth exploration of the PLP_deC gene family, particularly in the economically important Solanaceae family.

  19. The DACC system. Code burnup of cell for projection of the fuel elements in the power net work PWR and BWR

    International Nuclear Information System (INIS)

    Cepraga, D.; Boeriu, St.; Gheorghiu, E.; Cristian, I.; Patrulescu, I.; Cimporescu, D.; Ciuvica, P.; Velciu, E.

    1975-01-01

    The calculation system DACC-5 is a zero-dimensional reactor physics code used to calculate the criticality and burn-up of light-water reactors. The code requires as input essential extensive reactor parameters (fuel rod radius, water density, etc.). The nuclear constants (intensive parameters) are calculated with a five-group model (2 thermal and 3 fast groups). A fitting procedure is systematically employed to reduce the computation time of the code. Zero-dimensional burn-up calculations are made in an automatic way. Part one of the paper contains the code physical model and computer structure. Part two of the paper will contain tests of DACC-5 credibility for different light-water power lattices

  20. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    Energy Technology Data Exchange (ETDEWEB)

    Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

    2016-10-15

    Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

  1. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    International Nuclear Information System (INIS)

    Barkauskas, V.; Plukiene, R.; Plukis, A.

    2016-01-01

    Highlights: • RBMK-1500 fuel burn-up impact on k_e_f_f in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k_e_f_f in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k_e_f_f) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

  2. Construction and tests of a gamma device for experimental measurements of burnup of nuclear reactor fuel

    International Nuclear Information System (INIS)

    Brandao Junior, F.A.

    1982-01-01

    The gamma-scanning method is an important tool for the measurement of burnup of nuclear reactor fuel. The adequate knowledge of burnup allows for a better inventory of 'sensitive' fissile materials, better fuel management and provides insight on fuel behaviour and safety margins. This paper is related to the description, construction and operation of a first gamma scanning device, tested by irradiation of prototype PWR fuel pins, 14 cm long, in a Triga Mark-I reactor at very low power. Despite the limitations imposed by the low burnup, the experiment permitted a good checking of the main physical concepts and devices involved in the method. (Author) [pt

  3. SRAC-95, Cell Calculation with Burnup, Fuel Management for Thermal Reactors

    International Nuclear Information System (INIS)

    Tsuchihashi, K.; Ishiguro, Y.; Kaneko, K.; Ido, M.

    2004-01-01

    1 - Description of program or function: General neutronics calculation including cell calculation with burn-up, core calculation for any type of thermal reactor. Core burn-up calculation and fuel management by an auxiliary code. 2 - Method of solution: Collision probability method, 1D and 2D Sn for cell calculation; 1D, 2D and 3D diffusion for core calculation. 3 - Restrictions on the complexity of the problem: 20 regions for a continuous energy resonance absorption calculation and 16 steps for cell burn-up

  4. Burnup measurements on spent fuel elements of the RP-10 research reactor

    International Nuclear Information System (INIS)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro

    2011-01-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using 137 Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  5. Development of continuous energy Monte Carlo burn-up calculation code MVP-BURN

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Nakagawa, Masayuki; Sasaki, Makoto

    2001-01-01

    Burn-up calculations based on the continuous energy Monte Carlo method became possible by development of MVP-BURN. To confirm the reliably of MVP-BURN, it was applied to the two numerical benchmark problems; cell burn-up calculations for High Conversion LWR lattice and BWR lattice with burnable poison rods. Major burn-up parameters have shown good agreements with the results obtained by a deterministic code (SRAC95). Furthermore, spent fuel composition calculated by MVP-BURN was compared with measured one. Atomic number densities of major actinides at 34 GWd/t could be predicted within 10% accuracy. (author)

  6. Discharge Burnup Evaluation of Natural Uranium Loaded CANFLEX-43 Fuel Bundle

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Kim, Yong Hee; Kim, Won Young; Park, Joo Hwan

    2009-11-01

    Using WIMS-AECL code, which is 2-dimensional lattice core used in CANDU physics calculation, the discharge burnup of the natural uranium loaded CANFLEX-43 fuel bundle was evaluated by comparing the discharge burnup of standard 37 element fuel bundle. When the discharge burnup of the standard 37 element fuel is 7,200 MWd/MTU, that of the CANFLEX 43 fuel bundle was evaluated as 7,077 MWd/MTU, by applying the same lattice conditions for both fuel bundles

  7. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.; Lake, W.H.

    1989-01-01

    It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples represent many of the system features and alternatives already available for use in developing a broadly based criticality safety strategy for implementing burnup credit in the design and operation of spent fuel transport casks. 4 refs., 1 tab

  8. The radial distribution of plutonium in high burnup UO2 fuels

    International Nuclear Information System (INIS)

    Lassmann, K.; O'Carroll, C.; Laar, J. van de; Walker, C.T.

    1994-01-01

    A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21 000 and 64 000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions. (orig.)

  9. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  10. Approach for implementing burnup credit in high-capacity truck casks

    International Nuclear Information System (INIS)

    Boshoven, J.; Hopf, J.; Su, S.

    1991-01-01

    General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks in 1992. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for the higher enrichments. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA presents here an approach to burnup credit analysis to be included in the Safety Analysis Report for Packaging (SARP). 6 refs., 2 figs., 5 tabs

  11. High Burnup Fuel: Implications and Operational Experience. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2016-08-01

    This publication reports on the outcome of a technical meeting on high burnup fuel experience and economics, held in Buenos Aires, Argentina in 2013. The purpose of the meeting was to revisit and update the current operational experience and economic conditions associated with high burnup fuel. International experts with significant experience in experimental programmes on high burnup fuel discussed and evaluated physical limitations at pellet, cladding and structural component levels, with a wide focus including fabrication, core behaviour, transport and intermediate storage for most types of commercial nuclear power plants

  12. Criterion for burn-up conditions in gas-cooled cryogenic current leads

    International Nuclear Information System (INIS)

    Bejan, A.; Cluss, E.M. Jr.

    1976-01-01

    Superconducting magnets are energized through helium vapour-cooled cryogenic current leads operating at high ratios of current to mass flow. The high current operation where lead temperature, runaway, and eventual burn-up are likely to occur is investigated. A simple criterion for estimating the burn-up operation conditions (current, mass flow) for a given lead geometry (cross-sectional area, length, heat exchanger area) is presented. This article stresses the role played by the available heat exchanger area in avoiding burn-up at high ratios of current to mass flow. (author)

  13. The research on burnup characteristic of doping burnable poison in PWR

    International Nuclear Information System (INIS)

    Qiang Shenglong; Qin Dong; Chai Xiaoming; Yao Dong

    2014-01-01

    In PWR core design, burnable poisons are usually used for reactive compensation and power flatten. The choice of burnable poisons and how to match burnup would be the key-points for a long-life core design. We study the burnup character of doping burnable poisons (such as natural element, manual nuclide and soluble boron) in the PWR by the core burnup code MOI based on Monte Carlo method. The results show that Hf, Er and Eu doping burnable poison would be applicable for the nuclear design research on the long-life PWR core. (authors)

  14. Burn-up physics in a coupled Hammer-Technion/Cinder-2 system and ENDF/B-V aggregate fission product thermal cross section validation

    International Nuclear Information System (INIS)

    Santos, A. dos.

    1990-01-01

    The new methodology developed in this work has the following purposes: a) to implement a burnup capability into the HAMMER-TECHNION/9/computer code by using the CINDER-2/10/computer code to perform the transmutation analysis for the actinides and fission products; b) to implement a reduced version of the CINDER-2 fission product chain structure to treat explicity nearly 99% of all original CINDER-2 fission product absorption in a typical PWR unit cell; c) to treat the effect of the fission product neutron absorption in an unit cell in a multigroup basis; d) to develop a tentative validation procedure for the ENOF/C-V stable and long-lived fission product nuclear data based on the available experimental data/11-14/. The analysis will be performed by using the reduce chain in the coupled system CINDER-2 to generate the time dependent effective four group cross sections for actinides and fission products and CINDER-2 to perform the complete transmutation analysis with its built-in chain structure. (author)

  15. Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

    Energy Technology Data Exchange (ETDEWEB)

    Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2012-07-01

    The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now. The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)

  16. Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

    International Nuclear Information System (INIS)

    Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V.; Boulyga, S.F.; Becker, J.S.

    2005-01-01

    An analytical method is described for the estimation of uranium concentrations, of 235 U/ 238 U and 236 U/ 238 U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10 -9 g/g to 2.0 x 10 -6 g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing 235 U/ 238 U and 236 U/ 238 U isotope ratios and the average value amounted to 9.4±0.3 MWd/(kg U). (orig.)

  17. Effectiveness of group cognitive therapy about opium addict complications on attitude of adolescents with drug dependent parents

    Directory of Open Access Journals (Sweden)

    Kaveh Hojjat

    2015-12-01

    Full Text Available Background and Aim: Statistics show that 30% to 40 % of  opium addicted fathers’ children are prone to substance abuse in the future. The present study aimed at assessing the effectiveness of cognitive therapy approach  to attitude changing of adolescents with substance dependent fathers. Materials and Methods:  In this controlled. field-trail randomized study. .data collection tool was “attitude to addiction questionnaire”. The study population was all male students in the first grade of high school in Maneh - Samalghan city. . Six sessions of group cognitive therapy based on the effectiveness of drug side-effects in drug-addicted fathers’ adolescent children’s attitude were held. The above-mentioned questionnaire was filled out before and after intervention. The obtained data  was fed into SPSS software (V: 16 using. Independent t-test .and paired t-test were used for analysis and P<0.05 was taken as the significant level. Results:  There were no significant differences between the two groups in pre-test regarding their attitude about drug abuse (P=.20%. Mean score variance from pre-test to post-test in the intervention group decreased, but in the control group, it showed a slight increase. This means that the intervention reduced the positive attitude towards drugs, but the changes were not statistically significant (p=0.57. Besides, among ten factors decisive in an individual’s attitude about addiction, only group cognitive therapy  was able  to decrease mean points of an individual’s attitude about drug abuse .. Significantly (P = 0.04. Conclusion: It was found that group cognitive therapy education about opium  addict complicationsdidn`t have a significant effect on the attitude of the students with addicted fathers. Thus, a change of adolescents’ attitude requires more research.

  18. Fuel element burnup measurements for the equilibrium LEU silicide RSG GAS (MPR-30) core under a new fuel management strategy

    International Nuclear Information System (INIS)

    Pinem, Surian; Liem, Peng Hong; Sembiring, Tagor Malem; Surbakti, Tukiran

    2016-01-01

    Highlights: • Burnup measurement of fuel elements comprising the new equilibrium LEU silicide core of RSG GAS. • The burnup measurement method is based on a linear relationship between reactivity and burnup. • Burnup verification was conducted using an in-house, in-core fuel management code BATAN-FUEL. • A good agreement between the measured and calculated burnup was confirmed. • The new fuel management strategy was confirmed and validated. - Abstract: After the equilibrium LEU silicide core of RSG GAS was achieved, there was a strong need to validate the new fuel management strategy by measuring burnup of fuel elements comprising the core. Since the regulatory body had a great concern on the safety limit of the silicide fuel element burnup, amongst the 35 burnt fuel elements we selected 22 fuel elements with high burnup classes i.e. from 20 to 53% loss of U-235 (declared values) for the present measurements. The burnup measurement method was based on a linear relationship between reactivity and burnup where the measurements were conducted under subcritical conditions using two fission counters of the reactor startup channel. The measurement results were compared with the declared burnup evaluated by an in-house in-core fuel management code, BATAN-FUEL. A good agreement between the measured burnup values and the calculated ones was found within 8% uncertainties. Possible major sources of differences were identified, i.e. large statistical errors (i.e. low fission counters’ count rates), variation of initial U-235 loading per fuel element and accuracy of control rod indicators. The measured burnup of the 22 fuel elements provided the confirmation of the core burnup distribution planned for the equilibrium LEU silicide core under the new fuel management strategy.

  19. First-Principles Momentum Dependent Local Ansatz Approach to the Momentum Distribution Function in Iron-Group Transition Metals

    Science.gov (United States)

    Kakehashi, Yoshiro; Chandra, Sumal

    2017-03-01

    The momentum distribution function (MDF) bands of iron-group transition metals from Sc to Cu have been investigated on the basis of the first-principles momentum dependent local ansatz wavefunction method. It is found that the MDF for d electrons show a strong momentum dependence and a large deviation from the Fermi-Dirac distribution function along high-symmetry lines of the first Brillouin zone, while the sp electrons behave as independent electrons. In particular, the deviation in bcc Fe (fcc Ni) is shown to be enhanced by the narrow eg (t2g) bands with flat dispersion in the vicinity of the Fermi level. Mass enhancement factors (MEF) calculated from the jump on the Fermi surface are also shown to be momentum dependent. Large mass enhancements of Mn and Fe are found to be caused by spin fluctuations due to d electrons, while that for Ni is mainly caused by charge fluctuations. Calculated MEF are consistent with electronic specific heat data as well as recent angle resolved photoemission spectroscopy data.

  20. BURNY-SQUID, 2-D Burnup of UO2 and Mix UO2 PuO2 Fuel in X-Y or R-Z Geometry

    International Nuclear Information System (INIS)

    Rosa, I.; Zara, G.; Guidotti, R.

    1974-01-01

    1 - Nature of physical problem solved: - Multigroup neutron diffusion and burnup equations for two- to five- energy groups over a rectangular region of the x-y or r-z plane. - For a given geometry and initial enrichment, it calculates the two- to five- group flux distributions, the nuclides burnt in a time step t, and then the flux distribution again. This process is repeated until the maximum burn-up is reached. - Criticality search by uniform variation of a control isotope. - Solution of problems with fuel having different geometrical parameters, by means of super-compositions. - Recycle and restart options are available. - UO 2 and PUO 2 -UO 2 fuel can be handled. 2 - Method of solution: The zero-dimension burn-up program RIBOT-5 is coupled with the two-dimension program SQUID and alternately executed. The differential equations are solved by the difference method. 3 - Restrictions on the complexity of the problem: 200 maximum number of compositions 10,000 maximum number of mesh points 5 maximum Number of groups. 4 maximum number of super-compositions. Diagonal symmetry allowed

  1. Development of one-energy group, two-dimensional, frequency dependent detector adjoint function based on the nodal method

    International Nuclear Information System (INIS)

    Khericha, Soli T.

    2000-01-01

    One-energy group, two-dimensional computer code was developed to calculate the response of a detector to a vibrating absorber in a reactor core. A concept of local/global components, based on the frequency dependent detector adjoint function, and a nodalization technique were utilized. The frequency dependent detector adjoint functions presented by complex equations were expanded into real and imaginary parts. In the nodalization technique, the flux is expanded into polynomials about the center point of each node. The phase angle and the magnitude of the one-energy group detector adjoint function were calculated for a detector located in the center of a 200x200 cm reactor using a two-dimensional nodalization technique, the computer code EXTERMINATOR, and the analytical solution. The purpose of this research was to investigate the applicability of a polynomial nodal model technique to the calculations of the real and the imaginary parts of the detector adjoint function for one-energy group two-dimensional polynomial nodal model technique. From the results as discussed earlier, it is concluded that the nodal model technique can be used to calculate the detector adjoint function and the phase angle. Using the computer code developed for nodal model technique, the magnitude of one energy group frequency dependent detector adjoint function and the phase angle were calculated for the detector located in the center of a 200x200 cm homogenous reactor. The real part of the detector adjoint function was compared with the results obtained from the EXTERMINATOR computer code as well as the analytical solution based on a double sine series expansion using the classical Green's Function solution. The values were found to be less than 1% greater at 20 cm away from the source region and about 3% greater closer to the source compared to the values obtained from the analytical solution and the EXTERMINATOR code. The currents at the node interface matched within 1% of the average

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

    International Nuclear Information System (INIS)

    Gohar, Y.; Zhong, Z.; Talamo, A.

    2009-01-01

    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 ∼375 kW including the fission power of ∼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 electrons and the

  3. Development and benchmark verification of a parallelized Monte Carlo burnup calculation program MCBMPI

    International Nuclear Information System (INIS)

    Yang Wankui; Liu Yaoguang; Ma Jimin; Yang Xin; Wang Guanbo

    2014-01-01

    MCBMPI, a parallelized burnup calculation program, was developed. The program is modularized. Neutron transport calculation module employs the parallelized MCNP5 program MCNP5MPI, and burnup calculation module employs ORIGEN2, with the MPI parallel zone decomposition strategy. The program system only consists of MCNP5MPI and an interface subroutine. The interface subroutine achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, data exchanging with MCNP5MPI. Also, the program was verified with the Pressurized Water Reactor (PWR) cell burnup benchmark, the results showed that it's capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

  4. Fission gas release and fuel rod chemistry related to extended burnup

    International Nuclear Information System (INIS)

    1993-04-01

    The purpose of the meeting was to review the state of the art in fission gas release and fuel rod chemistry related to extended burnup. The meeting was held in a time when national and international programmes on water reactor fuel irradiated in experimental reactors were still ongoing or had reached their conclusion, and when lead test assemblies had reached high burnup in power reactors and been examined. At the same time, several out-of-pile experiments on high burnup fuel or with simulated fuel were being carried out. As a result, significant progress has been registered since the last meeting, particularly in the evaluation of fuel temperature, the degradation of the global thermal conductivity with burnup and in the understanding of the impact on fission gas release. Fifty five participants from 16 countries and one international organization attended the meeting. 28 papers were presented. A separate abstract was prepared for each of the papers. Refs, figs, tabs and photos

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

    International Nuclear Information System (INIS)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

    2000-10-01

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

  6. Conservatism in the actinide-only burnup credit for PWR spent nuclear fuel packages

    International Nuclear Information System (INIS)

    Lancaster, D.B.; Rahimi, M.; Thornton, J.

    1996-01-01

    In May 1995, the U.S. Department of Energy (DOE) submitted a topical report to the U.S. Nuclear Regulatory Commission (NRC) to gain actinide-only burnup credit for spent nuclear fuel (SNF) storage, transportation, or disposal packages. After approval of this topical report, DOE intends further submittals to the NRC to acquire additional burnup credit (e.g., the topical does not use fission products and is limited to only the first 100 yr of disposal). The NRC has responded to the topical with its preliminary questions. To aid in evaluation of the method, a review of the conservatism in the actinide-only burnup credit methodology was performed. An overview of the actinide-only burnup credit methodology is presented followed by a summary of the conservatism

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

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

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

  8. Proceedings of a workshop on the use of burnup credit in spent fuel transport casks

    International Nuclear Information System (INIS)

    Sanders, T.L.

    1989-10-01

    The Department of Energy sponsored a workshop on the use of burnup credit in the criticality design of spent fuel shipping casks on February 21 and 22, 1988. Twenty-five different presentations on many related topics were conducted, including the effects of burnup credit on the design and operation of spent fuel storage pools, casks and modules, and shipping casks; analysis and physics issues related to burnup credit; regulatory issues and criticality safety; economic incentives and risks associated with burnup credit; and methods for verifying spent fuel characteristics. An abbreviated version of the DOE workshop was repeated as a special session at the November 1988 American Nuclear Society Meeting in Washington, DC. Each of the invited speakers prepared detailed papers on his or her respective topic. The individual papers have been cataloged separately

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  10. Taking burnup credit for interim storage and transportation system for BWR fuels

    International Nuclear Information System (INIS)

    Yoshioka, Ken-ichi; Ando, Y.; Kumanomido, H.; Sasaki, T.; Mitsuhashi, I.; Ueda, M.

    2001-01-01

    In order to establish a realistic burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of several modules such as TGBLA, ORIGEN, CITATION, MCNP, and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. A compact measurement system for a fuel assembly has been being developed to meet requirements for the burnup determination, the neutron emission-rate evaluation, and the nuclear materials management. For a spent MOX fuel, a neutron emission rate measurement method has been being developed. The system consists of Cd-Te detectors and / or fission chambers. Some model calculations were carried out for the latest design BWR fuels. The effect of taking burnup credit for a transportation cask is shown. (authors)

  11. Implementation of burnup credit in spent fuel management systems. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-08-01

    The purpose of this Technical Committee Meeting was to explore the status of international activities related to the use of burnup credit for spent fuel applications. This was the second major meeting on the issues of burnup credit for spent fuel management systems held since the IAEA began to monitor the uses of burnup credit in spent fuel management systems in 1997. Burnup credit for wet and dry storage systems is needed in many Member States to allow for increased initial fuel enrichment, and to increase the storage capacity and thus to avoid the need for extensive modifications of the spent fuel management systems involved. This document contains 31 individual papers presented at the Meeting; each of the papers was indexed separately.

  12. Implementation of burnup credit in spent fuel management systems. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2001-08-01

    The purpose of this Technical Committee Meeting was to explore the status of international activities related to the use of burnup credit for spent fuel applications. This was the second major meeting on the issues of burnup credit for spent fuel management systems held since the IAEA began to monitor the uses of burnup credit in spent fuel management systems in 1997. Burnup credit for wet and dry storage systems is needed in many Member States to allow for increased initial fuel enrichment, and to increase the storage capacity and thus to avoid the need for extensive modifications of the spent fuel management systems involved. This document contains 31 individual papers presented at the Meeting; each of the papers was indexed separately

  13. Calculation of heat rating and burn-up for test fuel pins irradiated in DR 3

    International Nuclear Information System (INIS)

    Bagger, C.; Carlsen, H.; Hansen, K.

    1980-01-01

    A summary of the DR 3 reactor and HP1 rig design is given followed by a detailed description of the calculation procedure for obtaining linear heat rating and burn-up values of fuel pins irradiated in HP1 rigs. The calculations are carried out rather detailed, especially regarding features like end pellet contribution to power as a function of burn-up, gamma heat contributions, and evaluation of local values of heat rating and burn-up. Included in the report is also a description of the fast flux- and cladding temperature calculation techniques currently used. A good agreement between measured and calculated local burn-up values is found. This gives confidence to the detailed treatment of the data. (author)

  14. Fission gas and iodine release measured up to 15 GWd/t UO2 burnup

    International Nuclear Information System (INIS)

    Appelhans, A.D.

    1983-01-01

    A summary is presented of the measured release of xenon, krypton and iodine up to 15 GWd/t UO 2 burnup for fuel centerline temperatures ranging from 950 to 1800 K, at average linear heat ratings of 15 to 35 kW/m. The IFA-430 is composed of four 1.28-m-long fuel rods containing 10% enriched UO 2 pellet fuel. Two of the fuel rods are connected, top and bottom, to a gas flow system that permits the fission gases released from the fuel pellets to be swept out of the rods during irradiation and measured via gamma spectrometry. The release/burnup increased significantly between 10 and 15 GWd/t burnup. Fuel temperature did not change. Increased releases were due to physical changes in the fuel-surface area. Changes appeared to be due to higher power operation and burnup

  15. Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

    International Nuclear Information System (INIS)

    Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi

    2011-01-01

    The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.

  16. Experimental studies of spent fuel burn-up in WWR-SM reactor

    Energy Technology Data Exchange (ETDEWEB)

    Alikulov, Sh. A.; Baytelesov, S.A.; Boltaboev, A.F.; Kungurov, F.R. [Institute of Nuclear Physics, Ulughbek township, 100214, Tashkent (Uzbekistan); Menlove, H.O.; O’Connor, W. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Osmanov, B.S., E-mail: bari_osmanov@yahoo.com [Research Institute of Applied Physics, Vuzgorodok, 100174 Tashkent (Uzbekistan); Salikhbaev, U.S. [Institute of Nuclear Physics, Ulughbek township, 100214, Tashkent (Uzbekistan)

    2014-10-01

    Highlights: • Uranium burn-up measurement from {sup 137}Cs activity in spent reactor fuel. • Comparison to reference sample with known burn-up value (ratio method). • Cross-check of the approach with neutron-based measurement technique. - Abstract: The article reports the results of {sup 235}U burn-up measurements using {sup 137}Cs activity technique for 12 nuclear fuel assemblies of WWR-SM research reactor after 3-year cooling time. The discrepancy between the measured and the calculated burn-up values was about 3%. To increase the reliability of the data and for cross-check purposes, neutron measurement approach was also used. Average discrepancy between two methods was around 12%.

  17. The build-up and characterization of nuclear burn-up wave in a fast ...

    Indian Academy of Sciences (India)

    K V Anoop

    2018-02-07

    Feb 7, 2018 ... evaluating the quality of the wave by the researchers working in the field of nuclear burn-up wave build-up and propagation. Keywords. ... However, there are concerns relating to the nuclear safety, ... Simulation studies have.

  18. Inference for binomial probability based on dependent Bernoulli random variables with applications to meta-analysis and group level studies.

    Science.gov (United States)

    Bakbergenuly, Ilyas; Kulinskaya, Elena; Morgenthaler, Stephan

    2016-07-01

    We study bias arising as a result of nonlinear transformations of random variables in random or mixed effects models and its effect on inference in group-level studies or in meta-analysis. The findings are illustrated on the example of overdispersed binomial distributions, where we demonstrate considerable biases arising from standard log-odds and arcsine transformations of the estimated probability p̂, both for single-group studies and in combining results from several groups or studies in meta-analysis. Our simulations confirm that these biases are linear in ρ, for small values of ρ, the intracluster correlation coefficient. These biases do not depend on the sample sizes or the number of studies K in a meta-analysis and result in abysmal coverage of the combined effect for large K. We also propose bias-correction for the arcsine transformation. Our simulations demonstrate that this bias-correction works well for small values of the intraclass correlation. The methods are applied to two examples of meta-analyses of prevalence. © 2016 The Authors. Biometrical Journal Published by Wiley-VCH Verlag GmbH & Co. KGaA.

  19. Technical Issues in the development of high burnup and long cycle fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures

  20. Technical Issues in the development of high burnup and long cycle fuel pellets

    International Nuclear Information System (INIS)

    Kim, Dong Joo; Yang, Jae Ho; Oh, Jang Soo; Kim, Keon Sik; Rhee, Young Woo; Kim, Jong Hun; Nam, Ik Hui

    2012-01-01

    Over the last half century, a nuclear fuel cycle, a fuel discharged burnup and a uranium enrichment of the LWR (Light Water Reactor) fuel have continuously increased. It was the efforts to reduce the LWR fuel cycle cost, and to make reactor operation more efficiently. Improved fuel and reactor performance contribute further to the reduction and management efficiency of spent fuels. The primary incentive for operating nuclear reactor fuel to higher burnup and longer cycle is the economic benefits. The fuel cycle costs could be reduced by extending fuel discharged burnup and fuel cycle length. The higher discharged burnup can increase the energy production per unit fuel mass or fuel assembly. The longer fuel cycle can increase reactor operation flexibility and reduce the fuel changing operation and the spent fuel management burden. The margin to storage capacity limits would be also increased because high burnup and long cycle fuel reduces the mass of spent fuels. However, increment of fuel burnup and cycle length might result in the acceleration of material aging consisting fuel assembly. Then, the safety and integrity of nuclear fuel will be degraded. Therefore, to simultaneously enhance the safety and economics of the LWR fuel through the fuel burnup and cycle extension, it is indispensable to develop the innovative nuclear fuel material concepts and technologies which can overcome degradation of fuel safety. New fuel research project to extend fuel discharged burnup and cycle length has been launched in KAERI. Main subject is to develop innovative LWR fuel pellets which can provide required fuel performance and safety at extended fuel burnup and cycle length. In order to achieve the mission, we need to know that what the impediments are and how to break through current limit of fuel pellet properties. In this study, the technical issues related to fuel pellets at high burnup were surveyed and summarized. We have collected the technical issues in the literatures