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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. ZADOC, 2 Group Time-Dependent Burnup in X-Y Geometry with Fuel Management

    International Nuclear Information System (INIS)

    1 - Nature of physical problem solved: Two neutron group diffusion equations for a square mesh in x-y geometry are solved to yield a power distribution. Burnup for one time step is simulated by interpolation in a library of two-group cross sections which forms part of the problem data. It is assumed that the power distribution is invariant during one time step, at the end of which a re-calculation of flux and power follows automatically. Burnup proceeds in a succession of time steps and a number of fuel management options are available. 2 - Method of solution: Standard finite difference methods are used. 3 - Restrictions on the complexity of the problem: The programme is restricted to two neutron groups. Limits on the number of mesh points are as follows - IBM 7030 61 x 61; IBM 7090 32 x 32; see also AEEW - R.425

  3. Burnup-dependent effect of lattice-level homogenization and group condensation on calculated kinetics parameters for CANDU-type lattices

    International Nuclear Information System (INIS)

    Highlights: • CANDU-type-lattice kinetics parameters are calculated using different adjoint-weighting approximations at different burnups. • Fine-group space-dependent adjoint weighting is the most accurate method of calculating the kinetics parameters. • Two-group lattice-homogenized adjoint weighting overestimates the effective delayed-neutron fraction by approximately 5%. • Fine-group lattice-homogenized adjoint weighting overestimates the effective delayed neutron fraction only by approximately 2%. - Abstract: Modern analysis of nuclear reactor transients uses space-time reactor kinetics methods. In the Canadian nuclear industry, safety analysis calculations use almost exclusively the Improved Quasistatic (IQS) flux factorization method. The IQS method, like all methods based on flux factorization, relies on calculating effective point kinetics parameters, which dominate the time behavior of the flux, using adjoint-weighted integrals. The accuracy of the adjoint representation influences the accuracy of the effective kinetics parameters. Routine full core calculations are not performed using detailed models and transport theory, but rather using a cell-homogenized model and two-group diffusion theory. This work evaluates the effect of homogenization and group condensation at different burnups, for three fuel types: natural-uranium (NU) fuel, low-void reactivity (LVR) fuel and Advanced CANDU Reactor (ACR) fuel. Results show that the use of a two-group lattice-homogenized adjoint consistently overestimates the effective delayed neutron fraction by approximately 5% for all three fuel types and over a wide burnup range. The use of a two-group lattice-homogenized adjoint also introduces errors in the effective neutron generation time, but these are at most 1.3% (and their sign changes with burnup). Errors tend to vary with burnup by approximately 1% (of the individual parameter value). If a 69-group lattice-homogenized adjoint is used, the errors drop to

  4. Burnup-dependent cross section data for research reactors

    International Nuclear Information System (INIS)

    Studies currently in progress consider research and test reactors which commonly have burnups of 50 atom percent in 235-U and may reach as high a 70 atom percent. At these levels of burnup changes in cross-section data with burnup become significant. Some preliminary studies of these effects lead to the development of a modified version of REBUS-2 which supports changes in cross-section data with burnup. This version of REBUS-2 allows for changes in the cross-section data only at each time sub-interval in the problem, and these cross-section changes for capture and fission are based on a least squares polynomial fit as a function of burnup. In this paper an attempt is made to evaluate the importance of burnup dependent data for the various isotopes and/or groups, and to assess the accuracy of this method by comparing the REBUS-2 results with results obtained from PDQ-7. The 10 MW IAEA benchmark problem has been selected for this study. A description of the reactor and the XY model can be found in the IAEA Guidebook. The EPRI-CELL4 code was used to generate burnup dependent cross section data for use with both REBUS-2 and PDQ-7. Cross-section data were generated at 10 time steps to a burnup of approximately 50 atom percent in 235-U. The agreement between the PDQ-7 results and the REBUS-2 results with fitted burnup dependent cross-section data are quite good. Burnup dependent cross sections are essential for accurate estimates of cycle lengths and reactivities, and low order polynomial fits of capture and fission data for selected isotopes and energy groups can provide this capability

  5. Burnup dependent core neutronic analysis for PBMR

    International Nuclear Information System (INIS)

    The strategy for core neutronics modeling is based on SCALE4.4 code KENOV.a module that uses Monte Carlo calculational methods. The calculations are based on detailed unit cell and detailed core modeling. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and the fuel kernels in the pebble. The core is then modeled by placing these pebbles randomly throughout the core, yet not loosing track of any one of them. For the burnup model, a cyclic manner is adopted by coupling the KENOV.a and ORIGEN-S modules. Shifting down one slice at each discrete time step, and inserting fresh fuel from the top, this cyclic calculation model continues until equilibrium burnup cycle is achieved. (author)

  6. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.

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

    International Nuclear Information System (INIS)

    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, UO2 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 UO2 to higher oxides. The oxidation of UO2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO2 to UO2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO2.4 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 UO2 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 UO2 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)

  8. Considerations on burn-up dependent RIA and LOCA criteria

    International Nuclear Information System (INIS)

    For RIA transients, a fuel failure threshold has been derived and compared with recent experimental data relevant for BWR and PWR fuel. The threshold can be applied to HZP and CZP transients, account taken for the different initial enthalpy and for the lower ductility at cold conditions. It can also be used for non-zero power transients, provided that a term accounting for the initial power is incorporated. The proposed threshold predicts reasonably well the results obtained in the CABRI and NSRR tests when the different state of the cladding, i.e. ductile or brittle, is taken into account. Apart from some exceptions discussed in the paper, such as the effect of oxide spalling, one should consider ductile state for HZP conditions and brittle state for CZP conditions. The threshold applies equally well to UO2 and MOX fuel, but the database on MOX is limited. For LOCA transients, the cladding limit may decrease with burn-up due to cladding corrosion and hydrogen pick-up. A provisional criterion shows that the predicted burn-up effect is moderate or negligible if one uses the results obtained with actual high burn-up cladding. On the other hand, a large effect is predicted based on the results obtained with non-irradiated, pre-hydrided cladding specimens. There is a question however on as to whether these specimens can be representative for high burn-up material. The experimental evidence is still scarce and more data on high burn-up cladding is needed in order to arrive to firm conclusions. Most of the data currently available relates to Zr-4 cladding. The experiments made on ZIRLO and M5 cladding show that these alloys have a RIA and LOCA behaviour similar to or better than Zr-4. However, the data is limited, especially for LOCA conditions, where only un-irradiated specimens have been tested so far. (author)

  9. The dependence of the global neutronic parameters on the fuel burnup for CANDU SEU43 core

    Energy Technology Data Exchange (ETDEWEB)

    Balaceanu, V. [Institute for Nuclear Research, Pitesti (Romania); Pavelescu, M. [Academy of Romanian Scientists, Bucharest (Romania)

    2010-05-15

    In order to reduce the total fuel costs for the CANDU reactors, mainly by extending the fuel burnup limits, some fuel bundle concepts have been developed in different CANDU owner countries. Therefore, in our Institute the SEU43 (Slightly Enriched Uranium with 43 fuel elements) project was started in early '90s. The neutronic behavior analysis of the CANDU core with SEU43 fuel was an important step in our project design. The objective of this paper is to highline an analysis of the neutronic behavior of the CANDU SEU43 core with the fuel burnup. More exactly, the study refers to the dependence of some global neutronic parameters, mainly the reactivity, on the fuel burnup. Two types of CANDU core were taken into consideration: reference core (without any reactivity devices) and perturbed core (with a strong reactivity system inserted). The considered reactivity system is the Mechanical Control Absorber (MCA) one. The performed parameters are: k{sub eff.} values, the MCA reactivity worth and flux distributions. The fuel bundles in the core are SEU43, with the fuel enrichment in U{sup 235} of 0.96% and at nominal power. For the fuel burnup the values are: 0.00 GWd/tU (fresh fuel); 8.00 GWd/tU and 25.00 GWd/tU. For reaching this objective, a global neutronic calculation system named WIMSPIJXYZ LEGENTR is used. Starting from a 69-groups ENDF/B-V based library, this system uses three transport codes: (1) the standard lattice-cell code WIMS, for generating macroscopic cross sections in supercell option and also for burnup calculations; (2) the PIJXYZ code for 3D simulation of the MCA reactivity devices and the 3D correction of the macroscopic cross sections; (3) the LEGENTR 3D transport code for estimating global neutronic parameters (CANDU core). The analysis of the neutronic parameters consists of comparing the obtained results with the similar results calculated with the DRAGON and DIREN codes. This comparison shows a good agreement between these results. (orig.)

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

    OpenAIRE

    Muhammad Atta; Iqbal Masood; Mahmood Tayyab

    2011-01-01

    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 determin...

  11. 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.

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

    International Nuclear Information System (INIS)

    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. (author)

  13. Burn-Up Dependence of Bubble Morphology of Uranium Silicide Dispersion Fuels Used in Research Reactor

    International Nuclear Information System (INIS)

    Burn-up dependence of fission gas bubble morphology of U3Si2-Al and U3Si-Al dispersion fuels are reviewed with the data of ANL(Argonne Nation Laboratory) and KAERI(Korea Atomic Energy Research Institute

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

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

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

  15. Effect of burnup dependence of fuel cladding gap properties on WWER core characteristics

    International Nuclear Information System (INIS)

    Dependence of gas gap properties on burnup has been obtained with use of TRANSURANUS code. Implemented dependency on burnup is based on TRANSURANUS calculations of different fuel pins upon different linear power Ql. Obtained dependence was implemented into DYN3D code and results of new dependence effect on characteristics of WWER fuel loadings are presented. The work was performed in framework of orders BMU SR 2511 and BMU R0801504 (SR2611). The report describes the opinion and view of the contractor-State Scientific and Technical Centre on Nuclear and Radiation Safety-and does not necessarily represent the opinion of the ordering party-BMU-BfS/GRS and TUEV SUED. (Authors)

  16. 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...

  17. Verification of a Multi-group Cross Section Library for Burnup Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of); Joo, Hang Yu [Seoul National Univ., Seoul (Korea, Republic of)

    2013-05-15

    Despite satisfying the estimation of the neutronic parameters without depletion to some extent, it still requires detailed investigation of the behavior of a fuel with strong neutron absorber over its operating life time by nTRACER, the direct whole core calculation code with the conventional semi Predictor-Corrector method. This study is mainly focused on the verification of the newly generated multi-group library for burnup calculation by nTRACER through the analysis of its performance of depletion calculation of UO{sub 2} fuel with strong neutron absorbers such as Gadolinium. Firstly, the depletion calculation results of nTRACER are presented by comparing the evolution of k-inf and the inventories of commonly found important isotopes as a function of burnup in the cases of gadolinia(GAD)-bearing fuel pin and fuel assembly (FA) with those of MCNPX-version.2.6.0. The newly generated multi-group library for burnup calculation by nTRACER was verified through GAD-bearing fuel after the new approach of resonance treatment had been employed. Though very good agreement in the overall effect reflected on the multiplication factor of FA at BOC, the evolution of k-inf along fuel irradiation history was systematically well underestimated by nTRACER when compared to Monte Carlo results.

  18. The impact of time dependant spectra on fusion blanket burn-up

    International Nuclear Information System (INIS)

    Highlights: ► We modelled tritium production and nuclide burn-up within a spherical, solid-breeder blanket, with a 1 GW DT fusion source. ► The effect of updating reaction rates regularly is not significant for parent nuclides. ► Updating reaction rates regularly can change the daughter nuclide inventories by several hundred percent. ► Hydrogen and helium production within steels are not significantly effected by reaction rate update. ► A time step duration of 2 weeks or less is required for tritium breeding calculations. -- Abstract: Knowledge of nuclide burn-up within tritium breeding blankets has a crucial part to play in the safety, reliability and efficiency of fusion reactors. The modelling of burn-up requires a series of neutron transport calculations which can compute the reaction rate either directly, via Monte-Carlo estimators, or by implementing the multi-group method. These reaction rates can then be directly substituted into the burn-up equations, which can calculate nuclide number densities after a specified period of burn-up. The material burn-up will change the neutron spectra and the rate of nuclear reactions. Hence, a new neutron transport calculation needs to be performed after burn-up and the sequence is repeated for several time-steps. Radiation transport calculations are computationally expensive, therefore the minimisation of reaction rate calculations via Monte-Carlo simulations is desirable. Thus, time-intervals between Monte-Carlo simulations should be as large as possible. This paper addresses the effect of neutron spectra on the burn-up of parent and daughter nuclides found in EUROFER steel and the tritium self-sufficiency time. Using a spherical reactor geometry with lithium–lead tritium breeding material, a neutron spectrum is computed at time = 0 and time = 2 years after a detailed depletion calculation using 1 day time intervals. These two spectra are then used to calculate reaction rates for every isotope listed within

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

    International Nuclear Information System (INIS)

    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 U4O9r↓U3O8. Model results predict that U3O8 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

  20. Reactivity effects of nonuniform axial burnup distributions on spent fuel

    International Nuclear Information System (INIS)

    When conducting future criticality safety analyses on spent fuel shipping casks, burnup credit may play a significant role in determining the number of fuel assemblies that can be safely loaded into each cask. An important area in burnup credit analysis is the burnup variation along the length of the fuel assembly, which is determined by the location of the assembly in the reactor core and its residence time. A study of the effects of axial burnup distributions on reactivity has been conducted, using data from existing power plant fuel. Utilizing a one-dimensional, two-group diffusion code, named REALAX, the reactivity effects of axial burnup profiles have been calculated for various PWR fuel assemblies. The reactivity effects calculated by the code are defined in terms of k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup divided by k for a uniform axial burnup distribution at the assembly average burnup. Criticality safety specialists can take advantage of the quick-running code to determine axial effects of different assembly burnup profiles. In general, the positive reactivity effects of axial burnup distributions increase as burnup increases, though they do not increase faster than the overall decrease in reactivity due to burnup

  1. Reactivity effects of nonuniform axial burnup distributions on spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Leary, R.W. II; Parish, T.A. [Texas A & M Univ., College Station, TX (United States)

    1995-12-01

    When conducting future criticality safety analyses on spent fuel shipping casks, burnup credit may play a significant role in determining the number of fuel assemblies that can be safely loaded into each cask. An important area in burnup credit analysis is the burnup variation along the length of the fuel assembly, which is determined by the location of the assembly in the reactor core and its residence time. A study of the effects of axial burnup distributions on reactivity has been conducted, using data from existing power plant fuel. Utilizing a one-dimensional, two-group diffusion code, named REALAX, the reactivity effects of axial burnup profiles have been calculated for various PWR fuel assemblies. The reactivity effects calculated by the code are defined in terms of k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup divided by k for a uniform axial burnup distribution at the assembly average burnup. Criticality safety specialists can take advantage of the quick-running code to determine axial effects of different assembly burnup profiles. In general, the positive reactivity effects of axial burnup distributions increase as burnup increases, though they do not increase faster than the overall decrease in reactivity due to burnup.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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

  6. Effects of axial burnup distributions on the reactivity of spent fuel

    International Nuclear Information System (INIS)

    Criticality safety analyses for spent fuel shipping casks will eventually need to take credit for the decreased reactivity of spent fuel assemblies resulting from burnup. In order to do so, it will be necessary to assess the reactivity effects of the multitude of burnup shapes that can characterize spent fuel. A computer program, CASAX, has been written that allows the analyst to quickly evaluate the reactivity effects of actual and simplified axial burnup distributions on a group of PWR fuel assemblies. CASAX employs one dimensional, two group diffusion calculations to determine the k-effective of a cluster of assemblies. Assembly average, burnup dependent, two group cross sections for CASAX were obtained from CASMO3 using physical properties representative of Westinghouse 17 x 17 assemblies. Reactivity results are presented in terms of (k for the axially dependent burnup distribution minus k for a uniform axial burnup distribution at the assembly average burnup)/(k for a uniform axial burnup distribution at the assembly average burnup). Axial burnup distributions can have both positive and negative effects on the calculated k-effective. Positive reactivity effects generally result at high assembly average burnups and for axial distributions with low burnups in the assembly's tips

  7. Generation of one-group SELF shielded cross sections with multi-group approach for Monte Carlo burnup codes

    International Nuclear Information System (INIS)

    Allowing Monte Carlo (MC) codes to perform fuel cycle calculations requires coupling to a point depletion solver. In order to perform depletion calculations, one-group (1-g) cross sections must be provided in advance. This paper focuses on generating accurate 1-g cross section values that are necessary for evaluation of nuclide densities as a function of burnup. The proposed method is an alternative to the conventional direct reaction rate tally approach, which requires extensive computational efforts. The method presented here is based on the multi-group (MG) approach, in which pre-generated MG sets are collapsed with MC calculated flux. In our previous studies, we showed that generating accurate 1-g cross sections requires their tabulation against the background cross-section (σ0) to account for the self-shielding effect. However, in previous studies, the model that was used to calculate σ0 was simplified by fixing Bell and Dancoff factors. This work demonstrates that 1-g values calculated under the previous simplified model may not agree with the tallied values. Therefore, the original background cross section model was extended by implicitly accounting for the Dancoff and bell factors. The method developed here reconstructs the correct value of σ0 by utilizing statistical data generated within the MC transport calculation by default. The proposed method was implemented into BGCore code system. The 1-g cross section values generated by BGCore were compared with those tallied directly from the MCNP code. Very good agreement (<0.05%) in the 1-g cross values was observed. The method dose not carry any additional computational burden and it is universally applicable to the analysis of thermal as well as fast reactor systems. (author)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  10. Determination of dependence of fissile fraction in MOX fuels on spent fuel storage period for different burnup values

    International Nuclear Information System (INIS)

    Highlights: ► In a previous study, an expression to calculate fissile fraction of MOX for various burnups was obtained for 5-year cooled SF. ► In this follow-up study, a correction factor for spent fuel storage periods other than 5 years is derived. ► Thus, one major restriction on use of the expression derived in the initial study is eliminated. - Abstract: The purpose of this technical note is to remove one of the limitations of a derived expression in a previously published article (Özdemir et al., 2011). The original article focused on deriving (computationally) an expression for calculating total fissile fraction of mixed oxid (MOX) fuels depending on discharge burnup of spent fuel and desired burnup of MOX fuel; consequently, such an expression was obtained and put forward, together with its limitations. One of the limitations has been that all the computations and therefore the resulting expression are based on the assumption of a spent fuel storage period of 5 years. This follow-up study simply aims to obtain a correction factor for spent fuel storage periods other than 5 years; thus to remove one major restriction on use of the expression derived in the original article

  11. Development and validation of burnup dependent computational schemes for the analysis of assemblies with advanced lattice codes

    Science.gov (United States)

    Ramamoorthy, Karthikeyan

    The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant

  12. An advanced model for the prediction of the total burnup-dependent self-powered rhodium detector response

    International Nuclear Information System (INIS)

    This paper presents an advanced method to generate the burnup dependent total response of a rhodium self-powered detector operating in a pressurized water reactor environment. Full use is made of advanced nodal neutronic and coupled electron-photon transport techniques. The method accounts for (1) the detailed energy and spatial dependence of the neutron activation of each detector segment in a three-dimensional representation, (2) the generation of electrons caused by both neutron and gamma interactions in all the geometrical regions of the detector, and (3) the transport of the electrons within the detector to provide an observable current. All components of the detector signal are directly calculated - the method does not require the use of any empirical data, such as detector sensitivities. Intermediate results, such as beta escape fractions, were compared to measured data, and the overall technique was extensively benchmarked against operating data from three reactors

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

    International Nuclear Information System (INIS)

    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 UO2, UO2-Gd2O3, inhomogeneous MOX, and UO2-ThO2. 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 92U233-239, 93Np237-239, 94Pu238-243, 95Am241-244 (including isomers), and 96Cm242-245. Poisoning fission products are represented by 54Xe131,133,135, 48Cd113, 62Sm149,151,152, 64Gd154-160, 63Eu153,155, 36Kr83,85, 42Mo95, 43Tc99, 45Rh103, 47Ag109, 53I127,129,131, 55Cs133, 57La139, 59Pr141, 60Nd143-150, 61Pm147. Fission gases and volatiles included in the code are 36Kr83-86, 54Xe129-136, 52Te125-130, 53I127-131, 55Cs133-137, and 56Ba135-140. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

  14. The Mechanical Control Absorber reactivity dependence on the enrichment and burn-up in Candu-SEU super-cells

    Energy Technology Data Exchange (ETDEWEB)

    Balaceanu, Victoria; Constantin, Marin [Institute for Nuclear research, Pitesti (Romania)

    2008-07-01

    The objective of this work is to highlight some aspects of the local neutronic behaviour of the Candu SEU-43 fuel bundles (Slightly Enriched Uranium fuel bundles with 43 fuel elements). More exactly, the study refers to the dependence of some local neutronic parameters, mainly the reactivity, on the enrichment and the burn-up of the fuel. It was taken in consideration two types of super-cells: reference supercell (containing only fuel bundle and moderator) and perturbed supercell (containing fuel bundle, moderator and additionally a strong reactivity device). The considered reactivity device is the Mechanical Control Absorber (MCA). The performed parameters are: k{sub eff}. values, MCA reactivities and flux distributions. For reaching this objective, it is used a local neutronic calculation methodology based on WIMS and PIJXYZ codes. The paper ends with an analysis of the obtained results. (authors)

  15. 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)

  16. Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark

    International Nuclear Information System (INIS)

    Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.

  17. Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research; Panka, Istvan

    2015-09-15

    Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.

  18. Power ramped cladding stresses and strains in 3D simulations with burnup-dependent pellet–clad friction

    International Nuclear Information System (INIS)

    Highlights: ► This paper presents 2D plane strain and 3D simulations of pellet–cladding interaction during base irradiation and ramp tests. ► Inverse analysis is used to estimate the evolution of friction at the pellet–clad interface with burnup. ► The number of radial cracks that form in ramped rodlets is the main parameter on which inverse analysis is based. ► Calculations show that the sole evolution of the friction coefficient with burnup is sufficient to capture the radial crack pattern. ► A simple relation between the friction coefficient and the burnup is thus proposed and used in 3D simulations of PCI. - Abstract: This paper presents 2D(r, θ) plane strain and 3D simulations of PCI during base irradiation and ramp tests. Inverse analysis is used to estimate the evolution of friction at the pellet–clad interface with burnup. The number of radial cracks that form during power ramp tests in seventeen UO2-Zy4 rodlets with burnups in the range 20–60 GWd/tU is the main parameter on which inverse analysis is based. It is shown that the sole evolution of the friction coefficient with burnup is sufficient to capture the radial crack pattern of the rodlets after power ramping. A simple relation between the friction coefficient and the burnup variation after initial pellet–clad contact is thus proposed and used in 3D simulations of PCI. The delayed gap closing at mid-pellet level with respect to inter-pellet level leads to an axial variation of the friction coefficient, with maximum values near the pellet ends. The consequences in terms of PCI failure propensity are then discussed.

  19. Inclusion of historical dependences of fuel burn-up into MOBY-DICK Code

    International Nuclear Information System (INIS)

    The paper briefly describes inclusion of historical dependences of cross sections into MOBY-DICK code. Changes in program and its library are specified, especially from the point of view of programs user. Preliminary testing on assembly level is described and also testing on core level for 'transient' loading pattern calculations is demonstrated on examples of 18th and 19th cycles of the Dukovany NPP Unit III. Some features are addressed in the end of the paper (Authors)

  20. Fully Coupled Modeling of Burnup-Dependent (U1- y , Pu y )O2- x Mixed Oxide Fast Reactor Fuel Performance

    Science.gov (United States)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

    During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increase of fuel burnup, which decreases the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect the fuel material properties, power distribution, and overall performance of the fuel pin. In order to investigate these important issues, the (U1- y Pu y )O2- x fuel pellet is studied by fully coupling thermal transport, deformation, oxygen diffusion, fission gas release and swelling, and plutonium redistribution to evaluate the effects on each other with burnup-dependent models, accounting for the evolution of fuel porosity. The approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with fast reactor mixed oxide fuel performance analysis. The modeling results showed a consistent fuel performance comparable with the previous results. Burnup degrades the fuel thermal conductivity, resulting in a significant fuel temperature increase. The fission gas release increased rapidly first and then steadily with the burnup increase. The fuel porosity increased dramatically at the beginning of the burnup and then kept constant as the fission gas released to the fuel free volume, causing the fuel temperature to increase. Another important finding is that the deviation from stoichiometry of oxygen affects greatly not only the fuel properties, for example, thermal conductivity, but also the fuel performance, for example, temperature distribution, porosity evolution, grain size growth, fission gas release, deformation, and plutonium redistribution. Special attention needs to be paid to the deviation from stoichiometry of oxygen in fuel fabrication. Plutonium content will also affect the fuel material properties and performance

  1. 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)

    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

  2. Building neutron cross-section dependencies for few-group reactor calculations using stepwise regression

    International Nuclear Information System (INIS)

    Approximation of few-group neutron cross-sections by functions of burnup and thermal-hydraulics parameters of a fuel cell is considered. The cross-section is written as a sum of two terms: the base cross-section, which depends only on burnup and is computed under the nominal reactor core conditions, and the deviation, which depends on burnup and thermal-hydraulics variables of the cell. A one-dimensional dependence of the base cross-section is interpolated by a cubic spline. Multi-dimensional dependencies of the deviation are approximated by a polynomial. Construction of the polynomial is performed by a best-fitting selection of the polynomial terms using the stepwise regression algorithm. The number of terms to satisfy a user-given accuracy of approximation is minimized. As an example, approximation of a set of two-group macro and micro cross-sections as functions of burnup, coolant and fuel temperature, coolant density and boron concentration is considered for a fuel pin cell of a VVER reactor. The constructed five-dimensional polynomial approximating cross-sections within 0.05% tolerance has about 20 terms for fast group cross-sections and 50 terms for thermal group cross-sections. The error of approximation is verified on the two data sets: the initial data used for approximation and the test data being computed on randomly selected points. Mean square and maximum errors are comparable for all the cross-sections for both sets of data. These results show that the initial data can be applied to control the approximation error

  3. Instabilities of Monte-Carlo burnup calculations for nuclear reactors—Demonstration and dependence from time step model

    International Nuclear Information System (INIS)

    Graphical abstract: - Highlights: • Continuous Energy Monte-Carlo burnup code. • Instabilities of depletion calculation in loosely coupled system. • Advanced step model for burnup calculations. • Xenon profile oscillation in thermal reactor. • Parametrical study of instabilities. - Abstract: In this paper we use the Continuous Energy Monte-Carlo tool to expose the problem of burnup instabilities occurring in 1D and 2D systems based on PWR geometry. The intensity of power profile oscillations is studied as a function of geometry properties and time step length. We compare two step models for depletion procedure: classic staircase step model and stochastic implicit Euler method, that belongs to the family of predictor–corrector schemes. What is more, we consider the usage of better neutron source intensity value than beginning-of-step approximation. Required methodology was implemented into MCB5 simulation code. The practical conclusions about depletion calculations were formulated and the efficiency of advanced step model was confirmed

  4. Research on burnup physics

    International Nuclear Information System (INIS)

    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

  5. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

    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.)

  6. Stereological evolution of the rim structure in PWR-fuels at prolonged irradiation: Dependencies with burn-up and temperature

    International Nuclear Information System (INIS)

    The stereology of the rim-structure was studied for PWR-fuels up to the ninth irradiation cycle, achieving maximum local burn-ups of 240 GWd/tM and beyond. At intermediate radial positions (0.55 0 c = 0.29. Rim-cavities are expected to remain closed at least up to this limit

  7. 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...... engage in when participating in interdisciplinary research in a group, and we compare our findings with those of other studies in interdisciplinary research....

  8. Burnup span sensitivity analysis of different burnup coupling schemes

    International Nuclear Information System (INIS)

    Highlights: ► The objective of this work is the burnup span sensitivity analysis of different coupling schemes. ► Three kinds of schemes have been implemented in a new MCNP–ORIGEN linkage program. ► Two kinds of schemes are based predictor–corrector technique and the third is based on Euler explicit method. ► The analysis showed that the predictor–corrector approach better accounts for nonlinear behavior of burnup. ► It is sufficiently good to use the Euler method at small spans but for large spans use of second order scheme is mandatory. - Abstract: The analysis of core composition changes is complicated by the fact that the time and spatial variations in isotopic composition depend on the neutron flux distribution and vice versa. Fortunately, changes in core composition occur relatively slowly and hence the burnup analysis can be performed by dividing the burnup period into some burnup spans and assuming that the averaged flux and cross sections are constant during each burn up span. The burnup span sensitivity analysis attempts to find how much the burnup spans could be increased without any significant change in results. This goal has been achieved by developing a new MCNP–ORIGEN linkage program named MOBC (MCNP–ORIGEN Burnup Calculation). Three kinds of coupling scheme have been implemented in MOBC. Two of these are based on second order predictor–corrector technique and enable us to choose larger time steps, whilst the third one is based on Euler explicit first order method and is faster than the other two. The validity of the developed program has been evaluated by the code vs. code comparison technique. Two different types of codes are employed. The first one is based on deterministic two dimensional transport method, like CASMO-4 and HELIOS codes, and the second one is based on Monte Carlo method, like MCODE code. Only one coupling technique is employed in each of these state of the art codes, while the MOBC excels in its ability to

  9. Program package for 2D burnup calculation

    International Nuclear Information System (INIS)

    The program package for 2 dimension burnup calculation was developed for TRIGA Mark III reactor. The package consists of 3 modules: PRESIX, SIXTUS-2, and BURN; 1 library, and 2 input files. PRESIX module prepared cross sections for diffusion calculation. SIXTUS-2 module, a two dimensional diffusion code in hexagonal geometry, calculates keff, neutron fluxes and power distributions. BURN module performs the burnup of fuel elements and stored the result in the ELEM.DAT file. PRESIX.LIB is two group cross section library for major reactor core components prepared using WIMS-D4 code. PRES.INP, the first input file, reads information on reactor power and core loading pattern. ELEM.DAT, the second input file, is prepared for specific TRIGA reactor and dependent on operation history. To verify the reactor model and computational methods, the calculated excess reactivities were compared to the measurement. The results are in good agreement. (author)

  10. Burnup credit activities in the United States

    International Nuclear Information System (INIS)

    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)

  11. Fuel burnup analysis for the Moroccan TRIGA research reactor

    International Nuclear Information System (INIS)

    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

  12. Issues for effective implementation of burnup credit

    International Nuclear Information System (INIS)

    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)

  13. VVER-related burnup credit calculations

    International Nuclear Information System (INIS)

    The calculations related to a VVER burnup credit calculational benchmark proposed to the Eastern and Central European research community in collaboration with the OECD/NEA/NSC Burnup Credit Criticality Benchmark Working Group (working under WPNCS - Working Party on Nuclear Criticality Safety) are described. The results of a three-year effort by analysts from the Czech Republic, Finland, Germany, Hungary, Russia, Slovakia and the United Kingdom are summarized and commented on. (author)

  14. FTR tag burnup

    International Nuclear Information System (INIS)

    The gas tag burnup changes investigated were limited to the three tags (Kr-78/Kr-80, Xe-126/Xe-129 and Kr-82/Kr-80) currently accepted as being the most desirable. Control rod tag burnup was significantly greater than fuel rod tag burnup. This occurs because control rods stay in the reactor longer and occupy positions of greater low-energy flux. Thus, minimum tag spacings were set by the control rods as 1.079 for Kr-78/Kr-80, 1.189 for Xe-126/Xe-129 and 1.134 for Kr-82/Kr-80

  15. Burnup credit in Spain

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  17. The implementation of burnup credit in VVER-440 spent fuel

    International Nuclear Information System (INIS)

    The countries using Russian reactors VVER-440 cooperate in reactor physics in Atomic Energy Research (AER). One of topic areas is 'Physical Problems of Spent Fuel, Radwaste and Decommissioning' (Working Group E). In this article, in the first part is an overview about our activity for numerical and experimental verification of codes which participants use for calculation of criticality, isotopic concentration, activity, neutron and gamma sources and shielding is shown. The set of numerical benchmarks (CB1, CB2, CB3 and CB4) is very similar (the same idea, the VVER-440) to the OECD/NEA/NSC Burnup Credit Criticality Benchmarks, Phases 1 and 2. In the second part, verification of the SCALE 4.4 system (only criticality and nuclide concentrations) for VVER-440 fuel is shown. In the third part, dependence of criticality on burnup (only actinides and actinides + fission products) for transport cask C30 with VVER-440 fuel by optimal moderation is shown. In the last part, current status in implementation burnup credit in Slovakia is shown. (author)

  18. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

    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)

  19. COGEMA/TRANSNUCLEAIRE's experience with burnup credit

    International Nuclear Information System (INIS)

    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. Method of compensating distribution of reactor burnup degree

    International Nuclear Information System (INIS)

    An object of the present invention is to attain an appropriate power distribution and a burnup degree distribution during an operation cycle, thereby improving the succeeding operation cycle in a BWR type reactor. That is, a deviation between a distribution of an actual axial burnup degree and that of an aimed axial burnup degree in a reactor core is measured upon completion of the operation cycle by using a burnup degree distribution measuring device. Then, the content of burnable poisons in fresh fuels to be charged to the reactor core is controlled in accordance with the deviation, to compensate the distribution of the axial burnup degree in the reactor core in the next operation cycle. Accordingly, the distribution of the axial burnup degree in the reactor core can be made closer to the aimed distribution of the burnup degree in the next operation cycle. Further, appropriate power distribution and a burnup degree distribution can be obtained by improving the axial power distribution in the reactor core with the characteristics of the fresh fuels themselves to be loaded, without depending only on changes of a control rod pattern. Accordingly, fuel economy and operation performance can be improved. (I.S.)

  1. Detailed Burnup Calculations for Testing Nuclear Data

    Science.gov (United States)

    Leszczynski, F.

    2005-05-01

    A general method (MCQ) has been developed by introducing a microscopic burnup scheme that uses the Monte Carlo calculated fluxes and microscopic reaction rates of a complex system and a depletion code for burnup calculations as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy-dependent cross-section libraries and full 3D geometry of the system can be input for the calculations. The resulting predictions for the system at successive burnup time steps are thus based on a calculation route where both geometry and cross sections are accurately represented, without geometry simplifications and with continuous energy data, providing an independent approach for benchmarking other methods and nuclear data of actinides, fission products, and other burnable absorbers. The main advantage of this method over the classical deterministic methods currently used is that the MCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burnup codes adopted until now are the widely used MCNP and ORIGEN codes, but other codes can be used also. For using this method, there is need of a well-known set of nuclear data for isotopes involved in burnup chains, including burnable poisons, fission products, and actinides. For fixing the data to be included in this set, a study of the present status of nuclear data is performed, as part of the development of the MCQ method. This study begins with a review of the available cross-section data of isotopes involved in burnup chains for power and research nuclear reactors. The main data needs for burnup calculations are neutron cross sections, decay constants, branching ratios, fission energy, and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross

  2. Burnup and plutonium distribution of WWER-440 fuel pin at extended burnup

    International Nuclear Information System (INIS)

    The formation of rim region in LWR UO2 based nuclear fuel at high burnup is a common observation. This region has very high porosity due to excessive gas release. Such a region is also characterized by a significantly high plutonium concentration and high local burnup compared to the internal fuel region. Spatial distribution of these parameters has been incorporated with fuel behavior and performance analysis codes by using mostly empirical relations. Variation of these parameters depends on the neutron flux as well as neutron energy spectrum. Detailed neutronics analysis is necessary for the accurate prediction of these parameters. This study is performed by MCNP4B Monte Carlo code for the calculation of local neutron flux, ORIGEN2 for burnup and depletion calculations, and MONTEBURNS for coupling these codes. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell. Fuel pin is divided into a number of radial segments. A relatively small mesh size is used at the region near the surface to reveal the rim effect. The variation of plutonium and local burnup are obtained for high burnup. Results are compared with existing experimental observations for WWER-440 fuel and other theoretical predictions

  3. Application of a burnup verification meter to actinide-only burnup credit for spent PWR fuel

    International Nuclear Information System (INIS)

    A measurement system to verify reactor records for burnup of spent fuel at pressurized water reactors (PWR) has been developed by Sandia National Laboratories and tested at US nuclear utility sites. The system makes use of the Fork detector designed at Los Alamos National Laboratory for the safeguards program of the International Atomic Energy Agency. A single-point measurement of the neutrons and gamma- rays emitted from a PWR assembly is made at the center plane of the assembly while it is partially raised from its rack in the spent fuel pool. The objective of the measurements is to determine the variation in burnup assignments among a group of assemblies, and to identify anomalous assemblies that might adversely affect nuclear criticality safety. The measurements also provide an internal consistency check for reactor records of cooling time and initial enrichment. The burnup verification system has been proposed for qualifying spent fuel assemblies for loading into containers designed using burnup credit techniques. The system is incorporated in the US Department of Energy's.''Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages'' [DOE/RW 19951

  4. High burnup in DIONISIO code

    International Nuclear Information System (INIS)

    When the residence time of nuclear fuel rods exceeds a given threshold value, several properties of the pellet material suffer changes and hence the posterior behaviour of the rod is significantly altered. Structural modifications start at the pellet periphery, which is usually referred to as rim zone. It is presently believed that these changes are a consequence of the localized absorption of epithermal neutrons by 238U, which effective cross section presents resonant peaks. Due to the chain of nuclear reactions that take place, several Pu isotopes are born especially at the rim. In particular, the fissile character of 239Pu and 241Pu is the cause of the increased number of fission events that occur in the pellet periphery. For this reason, the power generation rate and the burnup adopt a non uniform distribution in the pellet, reaching at the rim values two or three times higher than the average [1]. The rim zone starts to form for a burnup threshold value of about 50-60 MWd/kgHM and its width increases as the irradiation progresses. The microstructure of this zone is characterized by the presence of small grains, with a typical size of 200 nm, and large pores, of some μm. Even though the rim zone is very thin, it has a significant effect on the mechanical integrity of the pellet, particularly when it makes contact with the cladding, and on the temperature distribution in the whole pellet, because of its low thermal conductivity [1,2]. The numerical codes designed to simulate fuel behaviour under irradiation must include the phenomena associated to high burnup if they aim at extending the prediction range, and this is the purpose with our DIONISIO code. But a detailed analysis of the phenomena that take place in this region demands the use of neutronic codes that solve the Boltzmann transport equations [3] in a number of energy intervals (groups), including adequate considerations in the region of the resonant absorption peaks of 238U. These cell codes predict

  5. On the dimensional dependence of the electromagnetic duality groups

    CERN Document Server

    Wotzasek, C

    1998-01-01

    We study the two-fold dimensional dependence of the electromagnetic duality groups. We introduce the dual projection operation that systematically discloses the presence of an internal space of potentials where the group operation is defined. A two-fold property of the kernel in the projection is shown to define the dimensional dependence of the duality groups. The dual projection is then generalized to reveal another hidden two-dimensional structure. The new unifying concept of the external duality space remove the dimensional dependence of the kernel, allowing the presence of both $Z_2$ and SO(2) duality groups in all even dimensions. This result, ultimately unifies the notion of selfduality to all D=2k+2 dimensions. Finally, we show the presence of an unexpected duality between the internal and external spaces leading to a duality of the duality groups.

  6. Prediction of fission gas release at high burn-up

    International Nuclear Information System (INIS)

    Reliable design of LWR fuel rods requires the fission gas release to be predicted as accurately as possible. Indeed that physical phenomenon governs both the fuel temperatures and the inner gas pressure. Fission gas release data have been reviewed by the NRC and it has been concluded that a fission gas release enhancement occurs at burn-up above 20 GWd/tM. To correct deficient fission gas release models which do not include burn-up dependence, the NRC developed an empirical correction method to describe burn-up enhancement effect. BELGONUCLEAIRE has developed its own fission gas release model which is utilized in licensing calculation through the COMETHE code. Fission gas release predictions at high burn-up are confronted to the experimental data as well as to the predictions of the NRC correlation. The physics of the fission gas release phenomenon is discussed

  7. SWAT, Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2

    International Nuclear Information System (INIS)

    1 - Description of program or function: SWAT evaluates isotopic composition of spent nuclear fuel, especially for burnup credit issues by driving codes SRAC95 and ORIGEN2.1 or ORIGEN2. SWAT is an automated driver code system. At the initial development phase, it was constructed by combining source programs of SRAC and ORIGEN2. To overcome the problem associated with code updates, SWAT chose to use system function of UNIX operating system to execute SRAC95 and ORIGEN2. So that, SWAT is independent of development and modification of SRAC95 and ORIGEN2.1. In SWAT, ORIGEN2(82) or ORIGEN2.1 is used for burnup calculations using the matrix exponential method. An updated decay library is included in the distribution. SWAT uses SRAC95 for neutron spectrum and effective cross section calculation in 107 groups, using the collision probability method for given geometry and isotopic composition. One or two dimensional cell geometries are supported in SRAC95. NEA-1698/02: The main purpose of new package is to run SWAT on several machines not supported in previous package (IA64 under Linux, Windows with cygwin and Sun,...) and several commercial FORTRAN compiler (Intel, PGI, Fujitsu). 2 - Methods: In calculating the problem-dependent cross section in SWAT, the total burnup history is divided into 'burnup steps'. Power, boric acid concentration, temperature of each region, and void ratio of coolant are given as history data. For each burnup step, the neutron spectrum and effective cross section are evaluated by SRAC95 using the information given in previous burnup calculation and cell geometry information. The user can select geometry options for the collision probability method in SRAC95. 3 - Restrictions on the complexity of the problem: Resonance absorption calculation with ultra-fine group cross section can not be directly applicable for 2D geometry

  8. Final evaluation of the CB3+burnup credit benchmark addition

    International Nuclear Information System (INIS)

    In 1966 a series of benchmarks focused on the application of burnup credit in WWER spent fuel management system was launched by L.Markova (1). The four phases of the proposed benchmark series corresponded to the phases of the Burnup Credit Criticality Benchmark organised by the OECD/NEA.These phases referred as CB1, CB2, CB3 and CB4 benchmarks were designed to investigate the main features of burnup credit in WWER spent fuel management systems. In the CB1 step, the multiplication factor of an infinite array of spent fuel rods was calculated taking the burnup, cooling time and different group of nuclides as parameters. The fuel compositions was given in the benchmark specification (Authors)

  9. Impact of extended burnup on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    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

  10. TRIGA criticality experiment for testing burn-up calculations

    Energy Technology Data Exchange (ETDEWEB)

    Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz [Jozef Stefan Institute, Reactor Physics Division, Ljubljana (Slovenia)

    1999-07-01

    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)

  11. A Burnup Analysis of PBMR-400MWth Reactor Core

    International Nuclear Information System (INIS)

    The purpose of this study is to analyze the burnup characteristics of 400MWth PBMR using Monte Carlo method. In the world, the deterministic method is widely used to model such that system but it still has a disadvantage which is not flexible in simulating the burnup cycle. Although this method applies some techniques to increase the accuracy of calculation results but it is necessary to model this system by a suitable computer code that can verify and validate the results of the deterministic method. A method which uses a Monte Carlo technique for simulating the burnup cycle was performed. A reactor physics computer code uses in this method is MONTEBURN 2.0 which accurately and efficiently computes the neutronic and material properties of the fuel cycle. MONTEBURN is a fully automated tool that links the MCNP Monte Carlo transport code with a radioactive decay and burnup code ORIGEN. In this model, the calculations are based on a detailed core modeling using MCNP. The fuel pebble is thoroughly modeled by introducing unit cell modeling for the graphite matrix and fuel kernels in the pebble. For the burnup model, a start-up core was studied with considering the movement of pebbles. By shifting down one layer at each discrete time step and inserting fresh fuel from the top, this cyclic calculation is continued until equilibrium burnup cycle is achieved. In this study, the time dependence of multiplication factor keff, the spatial dependence of flux profile, power distribution, burnup, and inventory of isotopes in the start up process are analyzed. The results will provide the basis data of the burnup process and be also utilized as the verified data to validate a compute code for PBMR core analysis which will be developed in near future

  12. High burnup experience in PWRs

    International Nuclear Information System (INIS)

    The purpose of this paper is to summarize the high burnup experience of Westinghouse PWR fuel. The emphasis is on two regions of commercial PWR fuel that attained region average burnups greater than 36,000 MWD/MTU. One region operated under load follow conditions. The other region operated at base load conditions with a high average linear heat rating. Coolant activity data and post irradiation data were obtained. The post-irradiation data consisted of visual examinations, crud sampling, rod-to-rod dimensional changes, fuel column length changes, rod and assembly growth, assembly bow, fuel rod profilometry, grid spring relaxation, and fuel assembly sipping tests. The data showed that the fuel operated reliably to this burnup. Plans for irradiation to higher burnups are also discussed

  13. Advanced Burnup Method using Inductively Coupled Plasma Mass Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Hilton, Bruce A. [Idaho Natonal Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Glagolenko, Irina; Giglio, Jeffrey J.; Cummings, Daniel G

    2009-06-15

    Nuclear fuel burnup is a key parameter used to assess irradiated fuel performance, to characterize the dependence of property changes due to irradiation, and to perform nuclear materials accountability. For advanced transmutation fuels and high burnup LWR fuels that have multiple fission sources, the existing Nd-148 ASTM burnup determination practice requires input of calculated fission fractions (identifying the specific fission source isotope and neutron energy that yielded fission, e.g., U-235 from thermal neutron, U-238 from fast neutron) from computational neutronics analysis in addition to the measured concentration of a single fission product isotope. We report a novel methodology of nuclear fuel burnup determination, which is completely independent of model predictions and reactor types. The proposed method leverages the capability of Inductively Coupled Plasma Mass Spectrometry (ICP-MS) to quantify multiple fission products and actinides and uses these data to develop a system of burnup equations whose solution is the fission fractions. The fission fractions are substituted back in the equations to determine burnup. This technique requires high fidelity fission yield data, which is not uniformly available for all fission products. We discuss different means that can potentially assist in indirect determination, verification and improvement (refinement) of the ambiguously known fission yields. A variety of irradiated fuel samples are characterized by ICP-MS and the results used to test the advanced burnup method. The samples include metallic alloy fuel irradiated in fast spectrum reactor (EBRII) and metallic alloy in a tailored spectrum and dispersion fuel in the thermal spectrum of the Advanced Test Reactor (ATR). The derived fission fractions and measured burnups are compared with calculated values predicted by neutronics models. (authors)

  14. Advanced Burnup Method using Inductively Coupled Plasma Mass Spectrometry

    International Nuclear Information System (INIS)

    Nuclear fuel burnup is a key parameter used to assess irradiated fuel performance, to characterize the dependence of property changes due to irradiation, and to perform nuclear materials accountability. For advanced transmutation fuels and high burnup LWR fuels that have multiple fission sources, the existing Nd-148 ASTM burnup determination practice requires input of calculated fission fractions (identifying the specific fission source isotope and neutron energy that yielded fission, e.g., U-235 from thermal neutron, U-238 from fast neutron) from computational neutronics analysis in addition to the measured concentration of a single fission product isotope. We report a novel methodology of nuclear fuel burnup determination, which is completely independent of model predictions and reactor types. The proposed method leverages the capability of Inductively Coupled Plasma Mass Spectrometry (ICP-MS) to quantify multiple fission products and actinides and uses these data to develop a system of burnup equations whose solution is the fission fractions. The fission fractions are substituted back in the equations to determine burnup. This technique requires high fidelity fission yield data, which is not uniformly available for all fission products. We discuss different means that can potentially assist in indirect determination, verification and improvement (refinement) of the ambiguously known fission yields. A variety of irradiated fuel samples are characterized by ICP-MS and the results used to test the advanced burnup method. The samples include metallic alloy fuel irradiated in fast spectrum reactor (EBRII) and metallic alloy in a tailored spectrum and dispersion fuel in the thermal spectrum of the Advanced Test Reactor (ATR). The derived fission fractions and measured burnups are compared with calculated values predicted by neutronics models. (authors)

  15. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    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)

  16. PWR AXIAL BURNUP PROFILE ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Acaglione

    2003-09-17

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

  17. Context dependent feature groups, a proposal for object representation.

    OpenAIRE

    Würtz, Rolf P.

    1997-01-01

    The usefulness of contextually guided processors is investigated a little further. A more general use for binding V1 cell responses than the one in the target article is proposed, which takes into account that strong responses of these cells can mean more than the presence of lines and edges. The possibility for different grouping depending on the activities of neighboring cells is essential for the approach.

  18. Triton burnup measurements by neutron activation at JT-60U

    International Nuclear Information System (INIS)

    This paper describes measurements on triton burnup in a deuterium plasma by the detection of the 2.5 MeV neutrons (from DD fusion) and the 14 MeV neutrons (from DT fusion). The 2.5 MeV neutrons have been measured by fission chambers and activation of indium foils while the 14 MeV neutrons have been detected by activation of silicon, aluminum, and copper foils. The measured yields of the 2.5 MeV neutrons utilizing In foils are similar 20-40% higher than the yields obtained from fission chambers depending on what calibration factors are used. The deviation decreases with the plasma major radius (or increasing plasma volume). When the triton burnup is measured by utilizing neutron threshold reactions (En>2.5 MeV) and In foils, then systematic errors in the calibration factors cancel and the maximum deviation between the measured triton burnup for different calibration factors is reduced to similar 5%. The measurements indicate that triton burnup increases with the 14 MeV neutron yield, indicating that the relative yield of 14 MeV neutrons increases depending on the time duration of the deuterium neutral beam injection (NBI). Furthermore, the triton burnup decreases with an increased plasma major radius, indicating increased triton ripple losses, and increases with plasma current, indicating reduced banana orbit losses. (orig.)

  19. Fuel burnup monitor for nuclear reactors

    International Nuclear Information System (INIS)

    An in-service detector is designed using the principle of comparing temperatures in the fuel element and in the detector material. The detector consists of 3 metallic heat conductors insulated with ceramic insulators, two of them with uranium fuel spheres at the end. One sphere is coated with zirconium, the other with zirconium and gold. The precision of measurement of the degree of fuel burnup depends on the precision of the measurement of temperature and is determined from the difference in temperature gradients of the two uranium fuel spheres in the detector. (M.D.)

  20. Fuel cycle cost considerations of increased discharge burnups

    International Nuclear Information System (INIS)

    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

  1. CANDU lattice uncertainties during burnup

    International Nuclear Information System (INIS)

    Uncertainties associated with fundamental nuclear data accompany evaluated nuclear data libraries in the form of covariance matrices. As nuclear data are important parameters in reactor physics calculations, any associated uncertainty causes a loss of confidence in the calculation results. The quantification of output uncertainties is necessary to adequately establish safety margins of nuclear facilities. In this work, microscopic cross-section has been propagated through lattice burnup calculations applied to a generic CANDU® model. It was found that substantial uncertainty emerges during burnup even when fission yield fraction and decay rate uncertainties are neglected. (author)

  2. Transnucleaire's experience with burnup credit in transport operations

    International Nuclear Information System (INIS)

    Facing a continued increase in fuel enrichment values, Transnucleaire has progressively implemented a burnup credit programme in order to maintain or, where possible, to improve the capacity of its transport packagings without physical modification. Many package design approvals, based on a notion of burnup credit, have been granted by the French competent authority for transport since the early eighties, and many of these approvals have been validated by foreign competent authorities. Up to now, these approvals are restricted to fuel assemblies made of enriched uranium and irradiated in pressurized water reactors (PWR). The characterization of the irradiated fuel and the reactivity of the package are evaluated by calculation, performed using qualified French codes developed by the CEA (Commisariat a l'Energie Atomique/French Atomic Energy Commission): CESAR as a depletion code and APOLO-MORET as a criticality code. The approvals are based on the hypothesis that the burnup considered is that applied on the least irradiated region of the fuel assemblies, the conservative approach being not to take credit for any axial profile of burnup along the fuel assembly. The most reactive configuration is calculated and the burnup credit is also restricted to major actinides only. On the operational side and in compliance with regulatory requirements, verification is made before transport, in order to meet safety objectives as required by the transport regulations. Besides a review of documentation related to the irradiation history of each fuel assembly, it consists of either a qualitative (go/no-go) verification or of a quantitative measurement, depending on the level of burnup credit. Thus the use of burnup credit is now a common practice with Transnucleaire's packages, particularly in France and Germany. New improvements are still in progress and qualifications of the calculation code are now well advanced, which will allow in the near future the use of six selected

  3. Determination of the accuracy of utility spent fuel burnup records. Interim report

    International Nuclear Information System (INIS)

    In order to develop a NRC-licensable burnup credit methodology, the pedigree and uncertainty of commercial spent nuclear fuel assembly burnup records needs to be established. Typically the assembly average burnup for each assembly is maintained in the plant records. It is anticipated that the repository for the disposal of spent fuel will utilize burnup credit and will require knowledge of the uncertainty of reactor burnup records. The uncertainty of the assembly average burnup record depends on the uncertainty of the method used to develop the record. Such records are generally based on core neutronic analysis coupled with analysis of in-core power detector data. This report evaluates the uncertainties in the burnup of fuel assemblies utilizing in-core measurements and core neutronic calculations for a Westinghouse PWR. To quantify the uncertainty, three cycles of in-core movable detector data were used. The data represents a first cycle of operation, a transition cycle and a low leakage cycle. These three cycles of data provide a true test of the uncertainty methodology. Three separate sets of results were used to characterize the burnup uncertainty of the fuel assemblies. The first set of results compared the measured and calculated reaction rates in instrumented assemblies and determined the uncertainty in the reaction rates. The second set of results determined the uncertainty in relative assembly power for both the instrumented and un-instrumented assemblies. The third set of results determined the burnup uncertainty of the discharged fuel in each cycle

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

    International Nuclear Information System (INIS)

    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)

  5. Burnup monitoring of VVER-440 spent fuel assemblies

    International Nuclear Information System (INIS)

    This paper reports on the results of the experiments performed on spent VVER-440 fuel assemblies at the Paks Nuclear Power Plant (NPP), Hungary. The fuel assemblies submerged in the service pit were examined by high-resolution gamma spectrometry (HRGS). The assemblies were moved to the front of a collimator tube built in the concrete wall of the pit in the reactor block at the NPP, and lifted down and up under water for scanning by the refueling machine. The HPGe detector was placed behind the collimator in an outside staircase. The measurements involved scanning of the assemblies along their length of all the 6 sides, at 5-12 measurement positions side by side. Axial and azimuthal burnup profiles were taken in this way. Assembly groups for measurements were selected according to their burnup (10–50 GWd/tU) and special positions (e. g. control assembly, neighbour of control assembly). Burnup differences were well observable between assembly sides looking towards the center of the core and opposite directions. Also, burnup profiles were different for control assemblies and normal (working) fuel assemblies. The ratio of the measured activities of Cs-134 and Cs-137 was evaluated by relative efficiency (intrinsic) calibration. Measurement uncertainty is around 3 %. Taking into account irradiation history and cooling time (i. e.the time elapsed since the discharge of the assembly out of the core), the activity ratio Cs-134/Cs-137 shows good correlation with the declared burnup.

  6. Development of high burnup fuel data-base

    International Nuclear Information System (INIS)

    Development of high burnup fuel data base (HBDB) was studied, which stores various performance data of high burnup fuels using a personal computer. Data items of the data base and storing and display methods of time-depending data such as power history were studied. It was shown that compound systems of a personal computer and an engineering work station have capacity for constructing the data base with much efficiency and small cost. And comparison of data items between the data base and the EPRI fuel base FPDB was discussed. (author)

  7. Perturbation and sensitivity theory for reactor burnup analysis

    International Nuclear Information System (INIS)

    Perturbation theory is developed for the nonlinear burnup equations describing the time-dependent behavior of the neutron and nuclide fields in a reactor core. General aspects of adjoint equations for nonlinear systems are first discussed and then various approximations to the burnup equations are rigorously derived and their areas for application presented. In particular, the concept of coupled neutron/nuclide fields (in which perturbations in either the neutron or nuclide field are allowed to influence the behavior of the other field) is contrasted to the uncoupled approximation

  8. CB2 result evaluation (VVER-440 burnup credit benchmark)

    International Nuclear Information System (INIS)

    The second portion of the four-piece international calculational benchmark on the VVER burnup credit (CB2) prepared in the collaboration with the OECD/NEA/NSC Burnup Credit Criticality Benchmarks Working Group and proposed to the AER research community has been evaluated. The evaluated results of calculations performed by analysts from Cuba, the Czech Republic, Finland, Germany, Russia, Slovakia and the United Kingdom are presented. The goal of this study is to compare isotopic concentrations calculated by the participants using various codes and libraries for depletion of the VVER-440 fuel pin cell. No measured values were available for the comparison. (author)

  9. Fast reactor 3D core and burnup analysis using VESTA

    Energy Technology Data Exchange (ETDEWEB)

    Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

  10. 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.

  11. Power excursion analysis for high burnup cores

    International Nuclear Information System (INIS)

    A study was undertaken of power excursions in high burnup cores. There were three objectives in this study. One was to identify boiling water reactor (BWR) and pressurized water reactor (PWR) transients in which there is significant energy deposition in the fuel. Another was to analyze the response of BWRs to the rod drop accident (RDA) and other transients in which there is a power excursion. The last objective was to investigate the sources of uncertainty in the RDA analysis. In a boiling water reactor, the events identified as having significant energy deposition in the fuel were a rod drop accident, a recirculation flow control failure, and the overpressure events; in a pressurized water reactor, they were a rod ejection accident and boron dilution events. The RDA analysis was done with RAMONA-4B, a computer code that models the space- dependent neutron kinetics throughout the core along with the thermal hydraulics in the core, vessel, and steamline. The results showed that the calculated maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important uncertainties in each of these categories are discussed in the report

  12. Economics of VVER Fuel Cycles Leading to High Discharge Burnup

    International Nuclear Information System (INIS)

    Economic characteristics of equilibrium VVER fuel cycles leading to high discharge burnup are investigated by supposing two scenarios named optimistic and pessimistic. The optimistic and pessimistic terms are used in the sense whether the high burnup fuel cycles are economically advantageous or the increasing enrichment cost can increase the specific fuel cycle cost above a certain discharge burnup value. Therefore in case of the optimistic scenario, maximum fabrication and back end costs and minimum enrichment and raw uranium costs were applied, while in case of the pessimistic scenario vice-versa. The applied costs are detailed in Table 1. Table1 Cost data of the two different scenarios. Concerning the transport and storage during the front end fuel cycle, it was assumed that application of burnable poison solves the criticality problems caused by the increased enrichment. By using the advantage of the burnup credit, the subcriticality of the spent fuel storage and transport devices can also be proved. Large reserve in the biological shielding is supposed. According to the above argumentation, fixed cost of the front and back end fuel cycle was used in the calculations, except the enrichment, but a 700 $/pin extra fabrication cost of the burnable poison was taken into account. Instead of fixed batch fraction, fixed cycle length was assumed which is advantageous for maximizing the discharge burnup and for minimizing the burnable poison extra cost but disadvantageous concerning the availability factor, which is constant in the given calculations. Beside the economic characteristics, the feasibility of the cycles are investigated from the point of view of the most important safety related parameters like reactivity coefficients and shut down margin. The figure below shows the burnup dependent fuel cycle cost for the above two scenarios. (author)

  13. Burnup determination of water reactor fuel

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency in consultation with the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The meeting was hosted by the Commission of the European Communities, at the Transuranium Research Laboratory, Joint Research Centre Karlsruhe, in the Federal Republic of Germany. This subject was dealt with for the first time by the IAEA. It was found to correspond adequately to this type of Specialist Meeting and to be suitable at a moment when the extension of burnup constitutes a major technical and economical issue in fuel technology. It was stressed that analysis of highly burnt fuels, mixed oxides and burnable absorber bearing fuels required extension of the experimental data base, to comply with the increasing demand for an improved fuel management, including better qualification of reactor physics codes. Twenty-seven participants from eleven countries plus two international organizations attended the Meeting. Twelve papers were given during three technical sessions, followed by a panel discussion which allowed to formulate the conclusions of the meeting and recommendations to the Agency. In addition, participants were invited to give an outline of their national programmes, related to Burnup Determination of Water Reactor Fuel. A separate abstract was prepared for each of these 12 papers. Refs, figs and tabs

  14. Determination of research reactor fuel burnup

    International Nuclear Information System (INIS)

    This report was prepared by a Consultants Group which met during 12-15 June 1989 at the Jozef Stefan Institute, Yugoslavia, and during 11-13 July 1990 at the IAEA Headquarters in Vienna, Austria, with subsequent contributions from the Consultants. The report is intended to provide information to research reactor operators and managers on the different, most commonly used methods of determining research reactor fuel burnup: 1) reactor physics calculations, 2) measurement of reactivity effects, and 3) gamma ray spectrometry. The advantages and disadvantages of each method are discussed. References are provided to assist the reactor operator planning to establish a programme for burnup determination of fuel. Destructive techniques are not included since such techniques are expensive, time consuming, and not normally performed by the reactor operators. In this report, TRIGA fuel elements are used in most examples to describe the methods. The same techniques however can be used for research reactors which use different types of fuel elements. 22 refs, 13 figs, 2 tabs

  15. Establishing a PWR burn-up library

    International Nuclear Information System (INIS)

    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 3000C/155 bar, 1900C/140 bar and 1000C/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.)

  16. Measurement techniques for verifying burnup

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, R.I. (Sandia National Lab., Albuquerque, NM (US)); Bierman, S.R. (Pacific Northwest Lab., Richland, WA (US))

    1992-05-01

    Measurements of the nuclear radiation from spent reactor fuel are being considered to qualify assemblies for loading into casks that will be used to transport spent fuel from utility sites to a federal storage facility. To ensure nuclear criticality safety, the casks are being designed to accept assemblies that meet restrictions as to burnup, initial enrichment and cooling time. This paper reports that measurements could be used to ensure that only fuel assemblies that meet the restrictions are selected for loading.

  17. Measurement techniques for verifying burnup

    International Nuclear Information System (INIS)

    Measurements of the nuclear radiation from spent reactor fuel are being considered to qualify assemblies for loading into casks that will be used to transport spent fuel from utility sites to a federal storage facility. To ensure nuclear criticality safety, the casks are being designed to accept assemblies that meet restrictions as to burnup, initial enrichment and cooling time. This paper reports that measurements could be used to ensure that only fuel assemblies that meet the restrictions are selected for loading

  18. TRIGA fuel burn-up calculations and its confirmation

    International Nuclear Information System (INIS)

    The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and its relevant burn-up of six selected Spent Fuel Elements SPE(s). High-resolution gamma-ray spectroscopy based non-destructive method is employed to measure spent fuel parameters. By the employment of this method, the axial distribution of Cesium-137 for six SPE(s) is measured, resulting in the axial burn-up profiles. Knowing the exact irradiation history and material isotopic inventory of an irradiated FE, six SPE(s) are selected for on-site gamma scanning using a special shielded scanning device developed at the ATI. This unique fuel inspection unit allows to scan each millimeter of the FE. For this purpose, each selected FE was transferred to the fuel inspection unit using the standard fuel transfer cask. Each FE was scanned at a scale of 1 cm of its active length and the Cs-137 activity was determined as proved burn-up indicator. The measuring system consists of a high-purity germanium detector (HPGe) together with suitable fast electronics and on-line PC data acquisition module. The absolute activity of each centimeter of the FE was measured and compared with reactor physics calculations. The ORIGEN2, a one-group depletion and radioactive decay computer code, was applied to calculate the activity of the Cs-137 and the burn-up of selected SPE. The deviation between calculations and measurements was in range from 0.82% to 12.64%.

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

    International Nuclear Information System (INIS)

    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-352 relates to the man-machine systems research part of the meeting and is in one volume, HPR-352 Volume I. The corresponding collection of papers in the fuel and materials research are given in two volumes, HPR-351 Volume I and HPR-351 Volume II. The overall programme of the Loen Enlarged Meeting covering the man-machine systems research is given in the following pages. The papers with

  20. 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 I

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  2. Future disposal burnup credit process and effort

    International Nuclear Information System (INIS)

    The United States Department of Energy's Office of Civilian Radioactive Waste Management has developed a risk-informed, performance based methodology for disposal criticality analyses. The methodology is documented in the Disposal Criticality Analysis Methodology Topical Report, YMP/TR-004Q (YMP 2000). The methodology includes taking credit for the burnup of irradiated commercial light water reactor fuel in criticality analyses, i.e., burnup credit. This paper summarizes the ongoing and planned future burnup credit activities associated with the methodology. (author)

  3. Impact of Integral Burnable Absorbers on PWR Burnup Credit Criticality Safety Analyses

    International Nuclear Information System (INIS)

    The concept of taking credit for the reduction in reactivity of burned or spent nuclear fuel (SNF) due to fuel burnup is commonly referred to as burnup credit. The reduction in reactivity that occurs with fuel burnup is due to the net reduction of fissile nuclide concentrations and the production of actinide and fission-product neutron absorbers. The change in the inventory of these nuclides with fuel burnup, and the consequent reduction in reactivity, is dependent upon the depletion environment. Therefore, the use of burnup credit necessitates consideration of all possible fuel operating conditions, including the use of integral burnable absorbers (IBAs). The Interim Staff Guidance on burnup credit [1] issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office recommends licensees restrict the use of burnup credit to assemblies that have not used burnable absorbers (e.g., IBAs or burnable poison rods, BPRs). This restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. The reason for this restriction is that the presence of burnable absorbers during depletion hardens the neutron spectrum, resulting in lower 235U depletion and higher production of fissile plutonium isotopes. Enhanced plutonium production has the effect of increasing the reactivity of the fuel at discharge and beyond. Consequently, an assembly exposed to burnable absorbers may have a slightly higher reactivity for a given burnup than an assembly that has not been exposed to burnable absorbers. This paper examines the effect of IBAs on reactivity for various designs and enrichment/poison loading combinations as a function of burnup. The effect of BPRs, which are typically removed during operation, is addressed elsewhere [2

  4. Phenomena and Parameters Important to Burnup Credit

    International Nuclear Information System (INIS)

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

  5. Phenomena and parameters important to burnup credit

    International Nuclear Information System (INIS)

    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)

  6. HAMCIND, Cell Burnup with Fission Products Poisoning

    International Nuclear Information System (INIS)

    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

  7. Non-destructive burnup determination of PWR spent fuel using Cs-134/Cs-137 and Eu-154/Cs-137

    International Nuclear Information System (INIS)

    Burnups for 36 points of five rods in the G23 assembly of Kori unit 1 have been determined on the basis of gamma-ray spectrometric measurement of two isotopic ratios, Cs-134/Cs-137 and Eu-154/Cs-137 in combination with the results calculated by the SCALE4.4 SAS2H module. Benchmarking of the SAS2H module has been done for the existing experimental data of Cs-13134, Cs-137 and Eu-154 isotopic compositions in PWR spent fuel. The gamma ray counts of two isotopic ratios have been corrected with their branching ratios, decay rates and energy dependent counting efficiencies in order to get true ratios. The energy dependent counting efficiencies have been determined as a quadratic equation based on the gamma ray counts for Cs-134 and Eu-154 at fourth energy points. Finally, burnups have been determined by putting true ratios of two isotopic ratios to their burnup-to-ratio fitting functions, respectively. Then the measured burnups have been compared with the declared burnup by the nuclear power plant. It is revealed that burnups determined from Cs-134/Cs-137 are agreeable with the declared burnups in most cases within about 12% error except a measuring point of C13, one of G23 fuel rods. In the case of Eu-154/Cs-137, the measured burnup is much lower than the declared burnup, which seems to be derived from system errors. (author)

  8. Extension of the TRANSURANUS burn-up model

    International Nuclear Information System (INIS)

    The validation range of the model in the TRANSURANUS fuel performance code for calculating the radial power density and burn-up in UO2 fuel has been extended from 64 MWd/kgHM up to 102 MWd/kgHM, thereby improving also its precision. In addition, the first verification of calculations with post-irradiation examination data is reported for LWR-MOX fuel with a rod average burn-up up to 45 MWd/kgHM. The extension covers the inclusion of new isotopes in order to account for the production of 238Pu. The corresponding one-group cross-sections used in the equations rely on results obtained with ALEPH, a new Monte Carlo burn-up code. The experimental verification is based on electron probe microanalysis (EPMA) and on secondary ion mass spectrometry (SIMS) as well as radiochemical data of fuel irradiated in commercial power plants. The deviations are quantified in terms of frequency distributions of the relative errors. The relative errors on the burn-up distributions in both fuel types remain below 12%, corresponding to the experimental scatter

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

    International Nuclear Information System (INIS)

    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)

  10. Determination of the burn-up of TRIGA fuel elements by calculation and reactivity experiments

    International Nuclear Information System (INIS)

    The burnup of 17 fuel elements of the TRIGA Mark-II reactor in Vienna was measured. Different types of fuel elements had been simultaneously used for several years. The measured burnup values are compared with those calculated on the basis of core configuration and reactor operation history records since the beginning of operation. A one-dimensional, two-group diffusion computer code TRIGAP was used for the calculations. Comparison with burnup values determined by γ-scanning is also made. (orig./HP)

  11. Development of high-burnup fuel analysis code EXBURN-I

    International Nuclear Information System (INIS)

    A computer code EXBURN-I has been developed which analyses LWR fuel behavior in high-burnup region in normal operation and transient conditions. In the high-burnup region, fuel behavior is affected considerably by such burnup-dependent factors as FP gas release, waterside corrosion of cladding, and pellet property change. To analyze these phenomena, in the present version, the base code FEMAXI-IV has been improved and incorporated such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding waterside corrosion. The present report describes the whole structure of the code, adopted models, and material properties, followed by input manual and sample input/output. Verification and further improvement of the code performance by experimental data will be done in the next stage. (author)

  12. Experimental programmes related to high burnup fuel

    International Nuclear Information System (INIS)

    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 LaC2, NdC2 and SmC2 have been measured by measuring the vapor pressures of CO over the region Ln2O3-LnC2-C during the carbothermic reduction of Ln2O3 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)

  13. The influence of pitch, burnup and absorber rods on the spent fuel pool criticality

    International Nuclear Information System (INIS)

    It has been shown that supercriticality might occur for some postulated accidents for the TRIGA spent fuel pool at ''Josef Stefan'' Institute in Ljubljana, Slovenia. However, in the previous studies, the effect of burnup was not accounted for. In this work the dependence of criticality on fuel burnup, the pitch among the elements and the number of uniformly mixed absorber rods for a square arrangement is presented. The Monte Carlo computer code MCNP4B with ENDF-B/VI library and detailed three dimensional geometry was used. WIMS-D code was used to model the isotopic composition of the fuel for 5, 10, 20 and 30 % burnup without cooling time. The results show, that out of the three studied effects: pitch from contact (3.75cm) up to rack design pitch (8cm), number of absorbers from 0 to 8 and burnup up to 30 %, the pitch has the greatest influence on the multiplication factor keff. In the interval in which the pitch was changed, keff decreased for up to 0.45. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g. for contact of fuel elements with 8 absorber rods among them, keff values are smaller for almost 0.20 than for arrangement without absorber rods regardless of the burnup. The effect of burnup is the smallest since in no case keff decreases for more than 0.10, even for high burnups of 30 %. (author)

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

    International Nuclear Information System (INIS)

    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

  15. High burnup fuel development program in Japan

    International Nuclear Information System (INIS)

    A step wise burnup extension program has been progressing in Japan to reduce the LWR fuel cycle cost. At present, the maximum assembly burnup limit of BWR 8 Χ 8 type fuel (B. Step II fuel) is 50GWd/t and a limited numbers of 9 Χ 9 type fuel (B. Step III fuel) with 55GWd/t maximum assembly burnup has been licensed by regulatory agencies recently. Though present maximum assembly burnup limit for PWR fuel is 48GWd/t (P. Step I fuel), the licensing work has been progressing for irradiation testing on a limited number of fuel assemblies with extended burnup of up to 55GWd/t (p. Step II fuel) Design of high burnup fuel and fabrication test are carried out by vendors, and subsequent irradiation test of fuel rods is conducted jointly by utilities and vendors to prepare for licensing. It is usual to make an irradiation test for vectarion, using lead use assemblies by government to confirm fuel integrity and reliability and win the public confidence. Nuclear Power Engineering Corporation (NUPE C) is responsible for verification test. The fuel are subjected to post irradiation examination (PIE) and no unfavorable indications of fuel behavior have found both in NUPE C verification test and joint irradiation test by utilities and vendors. Burnup extension is an urgent task for LWR fuel in Japan in order to establish the domestic fuel cycle. It is conducted in joint efforts of industries, government and institutes. However, watching a situation of burnup extension in the world, we are not going ahead of other countries in the achievement of burnup extension. It is due to a conservative policy in the nuclear safety of the country. This is the reason why the burnup extension program in Japan is progressing 'slow and steady' As for the data obtained, no unfavorable indications of fuel behavior have found both in NUPE C verification test and joint irradiation test by utilities and vendors until now

  16. 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.

  17. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

    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

  18. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    OpenAIRE

    M. H. Altaf; N.H. Badrun

    2014-01-01

    Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core) was found to remain as the hottest until 200 ...

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

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2002-10-23

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

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

    International Nuclear Information System (INIS)

    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

  1. FUMEX-III: A New IAEA Coordinated Research Project on Fuel Modelling at Extended Burnup

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency has initiated a new a Coordinated Research Project on Fuel Modelling at Extended Burnup (FUMEX-III). Currently, thirty one fuel modelling groups are participating with the intention of improving their capabilities to understand and predict the behaviour of water reactor fuel at high burnups. The exercise is carried in coordination with the OECD/NEA. The participants will model test cases provided by from sources such as the Halden Reactor Project and commercial irradiations and tests from the participants themselves. It is also intended to utilise idealised cases to test model behaviour under high burnup conditions. All cases are maintained in the OECD International Fuel Performance Experimental (IFPE) Database. The participants are particularly interested in modelling transient behaviour and mechanical interactions between pellet and cladding, including severe transient behaviour (RIA/LOCA) as well as temperature and fission gas release. However the participants include newcomer teams as well as state-of-the-art code users and have differing needs depending on the reactor system that they are modelling (PHWR, PWR, BWR, WWER) and the level of code development and experience that they have, so a matrix of test cases has been developed to allow each team to test their codes and methods appropriately. Some codes (eg TRANSURANUS and FEMAXI) are being used by several teams, both developing models and code user expertise. This paper summarises the objectives of the participants, the matrix of test cases that has been made available to the participants and some additional cases that are being prepared for inclusion in the later stages of the Project. (authors)

  2. Development of Inverse Estimation Program of Burnup Histories for Nuclear Spent Fuel Based on ORIGEN-S

    International Nuclear Information System (INIS)

    The purpose of this work is to develop a computer program which can accurately estimate burnup histories of spent fuels based on the environmental sample measurements. The burnup histories of spent fuels include initial uranium enrichment, discharge burnup, cooling time after discharge, and nuclear reactor type in which the spent fuel was burnt. The methodologies employed in our program are based on the formulations developed by M. R. Scott1 but we developed a stable bi-section method to correct initial uranium enrichment and used a simplified algorithm without burnup correction. Also, ORIGEN-S2 rather than ORIGEN-23 was used in our program to improve the accuracies by using the new capabilities of burnup dependent cross section libraries of ORIGEN-S. Our program is applied to several benchmark problems including realistic Mihama-3 problems to test the accuracies. We developed a computer program to determine the burnup history such as initial uranium enrichment, burnup, cooling time, and reactor type by using the results of sample measurements as input. Our methodologies are based on the methodologies given in Ref. 1 but we devised a new stable bisection method for the correction of initial uranium enrichment and we used ORIGEN-S rather than ORIGEN-2 to utilize the new capabilities of ORIGEN-S such as burnup dependent cross sections which can be prepared by using SCALE6

  3. Burnup credit issues in transportation and storage

    International Nuclear Information System (INIS)

    Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the US experience in developing a methodology to implement burnup credit in an integrated approach to transport cask design is presented in this paper. Specifically, technical issues related to the analysis, validation and implementation of burnup credit are identified and discussed

  4. Burnup credit issues in transportation and storage

    International Nuclear Information System (INIS)

    Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the U.S. experience in developing a methodology to implement burnup credit in an integrated approach to transport cask design is presented in this paper. Specifically, technical issues related to the analysis, validation and implementation of burnup credit are identified and discussed. (author)

  5. End effect analysis with various axial burnup distributions in high density spent fuel storage racks

    International Nuclear Information System (INIS)

    Highlights: • Criticality tests are carried out with various axial burnup distributions of fuel assemblies for spent fuel storage racks. • KENO-Va code system was used to obtain criticalities with 10 axial segments. • ORIGEN-S code system was used to obtain burnup dependent axial compositions. • The criticality and burnup dependent reactivity difference are obtained from the results. • End effect quantifications are satisfactory confirming the previous suggestions. - Abstract: End effect of spent fuel comes from the difference between uniform and actual axial burnup distributions of fuel assemblies. It is significant to control the criticality safety in spent fuel storage and transportation. This work is focused on estimation of end effect in the spent fuel of light water reactor for the spent fuel storage rack region-II. High and low burnups of corresponding different uranium enrichments are taken into consideration to analyze the end effect with different axial burnup distributions such as uniform, MOC and EOC profiles. Two types of fuel assemblies such as CE type and Westinghouse type are considered. The whole calculations have been carried out by using the SCALE6 code including ORIGEN-S and KENO-Va

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

    OpenAIRE

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

    2015-01-01

     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 u...

  7. Supercell burnup model for the physics design of BWR fuel assemblies

    International Nuclear Information System (INIS)

    A code called SUPERB has been developed for the BWR fuel assembly burnup analyses using supercell model. Each of the characteristic heterogeneities of a BWR fuel assembly like water gap, poisoned pins, control blade etc., is treated by invoking appropriate supercell concept. The burnup model of SUPERB is so devised as to strike a balance between accuracy and speed. This is achieved by building isotopic densities in each fuel pin separately while the depletion equations are solved only in a few groups of pins or burnup zones and the multigroup neutron spectra are differentiated in fewer group of pincell types. Multiple fuel ring burnup is considered only for Gd isotopes. A special empirical formula allows the microscopic cross section of Gd isotopes to be varied even during burnup integration. The supercell model has been tested against Monte Carlo results for the fresh cold clean Tarapur fuel assembly with two Gd fuel pins. The burnup model of SUPERB has been validated against one of the most sophisticated codes LWR-WIMS for a benchmark problem involving all the complexities of a BWR fuel assembly. The agreement of SUPERB results with both Monte Carlo and LWR-WIMS results is found to be excellent. (auth.)

  8. Addressing the Axial Burnup Distribution in PWR Burnup Credit Criticality Safety

    International Nuclear Information System (INIS)

    This paper summarizes efforts related to developing a technically justifiable approach for addressing the axial burnup distribution in PWR burnup-credit criticality safety analyses. The paper reviews available data on the axial variation in burnup and the effect of axial burnup profiles on reactivity in a SNF cask. A publicly available database of profiles is examined to identify profiles that maximize the neutron multiplication factor, keff, assess its adequacy for general PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. For this assessment, a statistical evaluation of the keff values associated with the profiles in the axial burnup profile database was performed that identifies the most reactive profiles as statistical outliers that are not representative of typical discharged SNF assemblies. The impact of these bounding profiles on the neutron multiplication factor for a high-density burnup credit cask is quantified. Finally, analyses are presented to quantify the potential reactivity consequence of assemblies with axial profiles that are not bounded by the existing database. The paper concludes with findings for addressing the axial burnup distribution in burnup credit analyses

  9. Extended burnup: fuel development and performance

    International Nuclear Information System (INIS)

    Fuel Performance for the B and W 15 x 15 (Mark B) and 17 x 17 (Mark C) fuel assembly designs is examined on a plant by plant basis. An extensive data base of fuel assembly and rod bow measurements and tests which demonstrate that these phenomena should not limit the high burnup capability of B and W fuel is presented. Post-irradiation measurements to date for fuel rod and assembly growth show that these phenomena are behaving as predicted and can be adequately evaluated and designed for in high burnup fuel assemblies. Clad creep and ductility data as a function of burnup for B and W fuel is presented with emphasis on their effects on our high burnup targets. Finally, fission gas release and waterside corrosion measurements results are presented

  10. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

    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)

  11. Determination of the fuel element burn-up for mixed TRIGA core by measurement and calculation with new TRIGLAV code

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, T.; Ravnik, M.; Persic, A. (J.Stefan Institute, Ljubljana (Slovenia))

    1999-12-15

    Results of fuel element burn-up determination by measurement and calculation are given. Fuel element burn-up was calculated with two different programs TRIGLAV and TRIGAC using different models. New TRIGLAV code is based on cylindrical, two-dimensional geometry with four group diffusion approximation. TRIGAC program uses one-dimensional cylindrical geometry with twogroup diffusion approximation. Fuel element burn-up was measured with reactivity method. In this paper comparison and analysis of these three methods is presented. Results calculated with TRIGLAV show considerably better alignment with measured values than results calculated with TRIGAC. Some two-dimensional effects in fuel element burn-up can be observed, for instance smaller standard fuel element burn-up in mixed core rings and control rod influence on nearby fuel elements. (orig.)

  12. REBUS: A burnup credit experimental programme

    International Nuclear Information System (INIS)

    An international programme called REBUS (REactivity tests for a direct evaluation of the Burn-Up credit on Selected irradiated LWR fuel bundles) for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Center SCK-CEN and Belgonucleaire. At present it is sponsored by USNRC, EdF from France and VGB, representing German nuclear utilities. The programme aims to establish a neutronic benchmark for reactor physics codes. This benchmark would qualify the codes to perform calculations of the burn-up credit. The benchmark exercise will investigate the following fuel types with associated burn-up. 1. Reference absorber test bundle, 2. Fresh commercial PWR UO2 fuel, 3. Irradiated commercial PWR UO2 fuel (50 GWd/tM), 4. Fresh PWR UO2 fuel, 5. Irradiated PWR UO2 fuel (30 GWd/tM). Reactivity effects will be measured in the critical facility VENUS. The accumulated burn-up of all rods will be measured non-destructively by gamma-spectrometry. Some rods will be analyzed destructively with respect to accumulated burn-up, actinides content and TOP-18 fission products (i.e. those non-gaseous fission products that have most implications on the reactivity). The experimental implementation of the programme will start in 2000. (author)

  13. Burnup credit implementation in spent fuel management

    International Nuclear Information System (INIS)

    The criticality safety analysis of spent fuel management systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. The concept of allowing reactivity credit for spent fuel offers economic incentives. Burnup Credit (BUC) could reduce mass limitation during dissolution of highly enriched PWR assemblies at the La Hague reprocessing plant. Furthermore, accounting for burnup credit enables the operator to avoid the use of Gd soluble poison in the dissolver for MOX assemblies. Analyses performed by DOE and its contractors have indicated that using BUC to maximize spent fuel transportation cask capacities is a justifiable concept that would result in public risk benefits and cost savings while fully maintaining criticality safety margins. In order to allow for Fission Products and Actinides in Criticality-Safety analyses, an extensive BUC experimental programme has been developed in France in the framework of the CEA-COGEMA collaboration. 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. Independent measurement systems, e.g. gamma spectrum detection systems, are needed to perform a true independent measurement of assembly burnup, without reliance on reactor records, using the gamma emission signatures fission products (mainly Cesium isotopes). (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  16. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science

    2014-12-15

    This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  17. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    International Nuclear Information System (INIS)

    This paper aims to investigate 232Th/233U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. 232Th/235U/238U oxide mixture was considered as fuel in the core, when the mass fraction of 232Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of 238U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the 232Th, 233U, 238U, 237Np, 239Pu, 241Am and 244Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  18. Activity ratio measurement and burnup analysis for high burnup PWR fuels

    International Nuclear Information System (INIS)

    Applying burnup credit to spent fuel transportation and storage system is beneficial. To take burnup credit to criticality safety design for a spent fuel transportation cask and storage rack, the burnup of target fuel assembly based on core management data must be confirmed by experimental methods. Activity ratio method, in which measured the ratio of the activity of a nuclide to that of another, is one of the ways to confirm burnup history. However, there is no public data of gamma-ray spectrum from high burnup fuels and validation of depletion calculation codes is not sufficient in the evaluation of the burnup or nuclide inventories. In this study, applicability evaluation of activity ratio method was carried out for high burnup fuel samples taken from PWR lead use assembly. In the gamma-ray measurement experiments, energy spectrum was taken in the Reactor Fuel Examination Facility (RFEF) of Japan Atomic Energy Agency (JAEA), and 134Cs/137Cs and 154Eu/137Cs activity ratio were obtained. With the MVP-BURN code, the activity ratios were calculated by depletion calculation tracing the operation history. As a result, 134Cs/137Cs and 154Eu/137Cs activity ratios for UO2 fuel samples show good agreements and the activity ratio method has good applicability to high burnup fuels. 154Eu/134Cs activity ratio for Gd2O3+UO2 fuels also shows good agreements between calculation results and experimental results as well as the activity ratio for UO2 fuels. It also becomes clear that it is necessary to pay attention to not only burnup but also axial burnup distribution history when confirming the burnup of UO2+Gd2O3 fuel with 134Cs/137Cs activity ratios. (author)

  19. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    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

  20. 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)

  1. The applications of burnup credit and the measurement techniques of burnup verification

    International Nuclear Information System (INIS)

    The factors of influencing criticality safety, implementing criticality control conditions, the calculation methods for predicting criticality, casks design and cask loading graph are described. The problems in the application of burnup credit and the dominant error in burnup credit operation are analysed. In order to avoid the operation error, requirements of measurement techniques and the most suitable measurement method are introduced

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

    International Nuclear Information System (INIS)

    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. 155Eu 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)

  3. 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)

  4. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    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

  5. Criticality safety analysis of WWER-440 spent fuel cask with radial and axial burnup profile implementation

    International Nuclear Information System (INIS)

    The impact of radial and axial burnup profile on the criticality of WWER-440 spent fuel cask is presented in the paper. The calculations are performed based on two AER Benchmark problems for WWER-440 irradiated fuel assembly. The radial zonewise dependent spent fuel inventory has been calculated by the NESSEL - NUKO code system. The axial dependent isotope concentrations have been determined by the modular code system SCALE4.4. For criticality calculations the SCALE4.4 has been applied. Calculations have been carried out for cask with 30 WWER-440 fuel assemblies with initial enrichment 3.6% of 235U and burnup up to 40 MWd/kgU. The influence of radial and axial burnup credit on the cask criticality has been evaluated

  6. Application of scale-4 depletion/criticality sequences in burnup credit analyses

    International Nuclear Information System (INIS)

    The concept of allowing reactivity credit for the transmuted state of spent fuel complicates the criticality analysis by requiring the specification of the fuel mixture to potentially include large numbers of isotopes representative of the fuel conditions. These conditions include the initial enrichment, local or average burnup conditions depending on the analysis approach, and the post-shutdown cooling time. In the development of an analysis methodology to evaluate spent fuel shipping and transport casks (flasks) based on this burnup credit, commercial reactor critical configurations were evaluated as potential experimental spent fuel criticals. This paper describes how the SCALE-4 depletion sequences (SAS2H), the cross-section processing sequence (CSASN), and the criticality module (KENO V.a) were used to evaluate these reactor criticals. A description of a newly developed sequence for linking SAS2H calculated burnup-dependent isotopics to KENO V.a mixing tables [SAS2H Nuclide Inventories for KENO Runs (SNIKR)] is also included

  7. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    International Nuclear Information System (INIS)

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

  8. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    Energy Technology Data Exchange (ETDEWEB)

    A.H. Wells

    2004-11-17

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

  9. Burnup analysis of the power reactor, 3

    International Nuclear Information System (INIS)

    The atomic number densities of uranium and transuranium were measured for JPDR-1. For the purpose of the study, the program has been prepared. It solves the burnup equation by the exponential matrix method. The void fraction and exposure distribution of the required data were calculated by three-dimensional nuclear-thermal-hydro-dynamic program FLORA under the operating conditions. The distribution of each atomic number density was obtained. The results agree with the measured values. The programs calculating nuclear constants in the cell were evaluated by obtaining the effective cross sections from the atomic number densities and the burnup. (auth.)

  10. Fission gas release modelling at high burnup

    International Nuclear Information System (INIS)

    A large quantity of experimental data on fission gas release is now available in the public domain. It covers a wide variety of fuel types and burnups of up to more than 70 GWd/tU. This data, together with gas release measurements from British Energy's AGRs, has been used to build a comprehensive validation database for the fuel performance code ENIGMA. Validation of ENIGMA version 5.11 against this database has identified a requirement for model development to improve predictions at high burnup. A modified gas release model has been produced and tested. (author)

  11. Nuclear fuel burn-up economy

    International Nuclear Information System (INIS)

    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)

  12. Methods of RECORD, an LWR fuel assembly burnup code

    International Nuclear Information System (INIS)

    The RECORD computer code is a detailed rector physics code for performing efficient LWR fuel assembly calculations, taking into account most of the features found in BWR and PWR fuel designs. The code calculates neutron spectrum, reaction rates and reactivity as a function of fuel burnup, and it generates the few-group data required for use in full scale core simulation and fuel management calculations. The report describes the methods of the RECORD computer code and the basis for fundamental models selected, and gives a review of code qualifications against measured data. (Auth. /RF)

  13. OTTER 3 - A single channel, axial burnup code

    International Nuclear Information System (INIS)

    OTTER 3 is a single channel, axial burnup code, written in Fortran for the KDF 9 computer, and suitable for studying fuel management schemes of the continuous charge/discharge type. A general fuel shuffling scheme is allowed, and both unidirectional and bidirectional fuel feed can be studied. A 2-group neutron diffusion code is incorporated, the flux equations being solved by the forward elimination - backward substitution technique for the inner problem and a source iteration technique accelerated by Chebyshev extrapolation for the outer problem. (author)

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

    International Nuclear Information System (INIS)

    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

  15. BISON, 1-D Burnup and Transport in Slab, Cylindrical, Spherical Geometry

    International Nuclear Information System (INIS)

    1 - Description of problem or function: BISON-1.5 solves the one- dimensional Boltzmann transport equation for neutron and gamma-rays and transmutation equations for fuel nuclides. 2 - Method of solution: In the transport calculation stage the one- dimensional Boltzmann transport equation is solved by the discrete ordinates method. In the burnup calculation stage, transmutation equations for fuel nuclides are solved by Bateman's method. The neutron flux obtained in the transport calculation stage is used to determine the transmutation rates in the burnup calculation stage. Both stages are repeated in tandem till the end of the burnup cycle. 3 - Restrictions on the complexity of the problem: A 42-group neutron and 21-group gamma-ray cross section library is prepared in the code package. Core storage for array variables is dynamically allocated by the code, so there are no restrictions on the size of each array

  16. NFCSim: A Dynamic Fuel Burnup and Fuel Cycle Simulation Tool

    International Nuclear Information System (INIS)

    NFCSim is an event-driven, time-dependent simulation code modeling the flow of materials through the nuclear fuel cycle. NFCSim tracks mass flow at the level of discrete reactor fuel charges/discharges and logs the history of nuclear material as it progresses through a detailed series of processes and facilities, generating life-cycle material balances for any number of reactors. NFCSim is an ideal tool for analysis - of the economics, sustainability, or proliferation resistance - of nonequilibrium, interacting, or evolving reactor fleets. The software couples with a criticality and burnup engine, LACE (Los Alamos Criticality Engine). LACE implements a piecewise-linear, reactor-specific reactivity model for its criticality calculations. This model constructs fluence-dependent reactivity traces for any facility; it is designed to address nuclear economies in which either a steady state is never obtained or is a poor approximation. LACE operates in transient and equilibrium fuel management regimes at the refueling batch level, derives reactor- and cycle-dependent initial fuel compositions, and invokes ORIGEN2.x to carry out burnup calculations

  17. Studies on future application of burnup credit in Hungary

    International Nuclear Information System (INIS)

    This paper describes the present status of the fuel storage and the possible future applications of burnup credit in wet and dry storage systems in Hungary. It gives a survey of the activities planned in AERI concerning the burnup credit. Some part of these investigations dealing with the influence of the axial changing of the assembly burnup are given in more details. (author)

  18. Transient behaviour of high burnup fuel

    International Nuclear Information System (INIS)

    The main subjects of the meeting were the discussion of regulatory background, integral tests and analysis, plant calculations, separate-effect test and analysis, concerning high burnup phenomena during RIA accidents in reactors, especially LWR, BWR and PWR type reactors. 32 papers were abstracted and indexed individually for the INIS database. (R.P.)

  19. Optimum burnup of BAEC TRIGA research reactor

    International Nuclear Information System (INIS)

    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

  20. Benchmark calculation with MOSRA-SRAC for burnup of a BWR fuel assembly

    International Nuclear Information System (INIS)

    The Japan Atomic Energy Agency has developed the Modular Reactor Analysis Code System MOSRA to improve the applicability of neutronic characteristics modeling. The cell calculation module MOSRA-SRAC is based on the collision probability method and is one of the core modules of the MOSRA system. To test the module on a real-world problem, it was combined with the benchmark program 'Burnup Credit Criticality Benchmark Phase IIIC.' In this program participants are requested to submit the neutronic characteristics of burnup calculations for a BWR fuel assembly containing fuel rods poisoned with gadolinium (Gd2O3), which is similar to the fuel assembly at TEPCO's Fukushima Daiichi Nuclear Power Station. Because of certain restrictions of the MOSRA-SRAC burnup calculations part of the geometry model was homogenized. In order to verify the validity of MOSRA-SRAC, including the effects of the homogenization, the calculated burnup dependent infinite multiplication factor and the nuclide compositions were compared with those obtained with the burnup calculation code MVP-BURN which had already been validated for many benchmark problems. As a result of the comparisons, the applicability of MOSRA-SRAC module for the BWR assembly has been verified. Furthermore, it can be shown that the effects of the homogenization are smaller than the effects due to the calculation method for both multiplication factor and compositions. (author)

  1. Burn-up and cycle length optimization project of the robust fuel programme

    International Nuclear Information System (INIS)

    The Spanish electric sector (UNESA) takes part in the Robust Fuel programme in the different work groups set up by EPRI. Iberinco, with the collaboration of Iberdrola Generacion (TECNO and Cofrentes NPP) and Soluziona Ingenieria, has created a stable multidisciplinary group to assimilate and follow up this program, analyzing in detail the technology generated and evaluating the conclusions to provide the most suitable recommendations for application. Along these lines, one of the most promising projects within technical group 3 (High burn properties) has been the one called Burn-up and cycle Length Optimization. In January 2000 Duke Power published a study on the plants it owns (PWR type) and 18-month cycles, to establish the optimum unloading burn-up of fuel. The conclusion it reached is that the fuel cost drops t a minimum for average unload burn-ups of between 60 and 70 GWd/MTU. As an extension to this study and covering a wider base of considerations, Exelon, with the support of Westinghouse and the University of Pennsylvania, released a study in December 2001 on different reference cores with different cycle lengths. In this study, the optimum burn-up without exceeding current maximum enrichment limits (5%) is determined. Publication of the results of the second phase, considering higher enrichments, was due in the summer of 2002. The design of the core to be refueled and economic analyzes show that both pressurized water reactors (PWR) and boiling water reactors (BWR) can obtain significant benefits by increasing the fuel unloading burn-up above currently licensed limits. However, the optimum unload burn-up level is not reached without exceeding the current enrichment limit of 5% . (Author)

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

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    uses an empirical gas release model combined with a strongly burn-up dependent correction term, developed by the US Nuclear Regulatory Commission. The paper presents the experimental results and the code calculations. It is concluded that the model predictions are in reasonable agreement (within 15...

  3. Development of burnup calculation function in reactor Monte Carlo code RMC

    International Nuclear Information System (INIS)

    This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua University of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including the middle-of-step approximation and the predictor-corrector method, are adopted by RMC to assure the accuracy under large burnup step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably saves computational time with negligible accuracy loss. According to the validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency. (authors)

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

    International Nuclear Information System (INIS)

    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

  5. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ozdemir, Levent, E-mail: levent.ozdemir@taek.gov.tr [Department of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara (Turkey); Acar, Banu Bulut; Zabunoglu, Okan H. [Department of Nuclear Engineering, Hacettepe University, 06800 Beytepe, Ankara (Turkey)

    2011-02-15

    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 {sup 239}Pu and {sup 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.

  7. Assessing Colour-dependent Occupation Statistics Inferred from Galaxy Group Catalogues

    CERN Document Server

    Campbell, Duncan; Hearin, Andrew; Padmanabhan, Nikhil; Berlind, Andreas; Mo, H J; Tinker, Jeremy; Yang, Xiaohu

    2015-01-01

    We investigate the ability of current implementations of galaxy group finders to recover colour-dependent halo occupation statistics. To test the fidelity of group catalogue inferred statistics, we run three different group finders used in the literature over a mock that includes galaxy colours in a realistic manner. Overall, the resulting mock group catalogues are remarkably similar, and most colour-dependent statistics are recovered with reasonable accuracy. However, it is also clear that certain systematic errors arise as a consequence of correlated errors in group membership determination, central/satellite designation, and halo mass assignment. We introduce a new statistic, the halo transition probability (HTP), which captures the combined impact of all these errors. As a rule of thumb, errors tend to equalize the properties of distinct galaxy populations (i.e. red vs. blue galaxies or centrals vs. satellites), and to result in inferred occupation statistics that are more accurate for red galaxies than f...

  8. High-Resolution Crystal Structures Elucidate the Molecular Basis of Cholera Blood Group Dependence

    Science.gov (United States)

    Heggelund, Julie Elisabeth; Burschowsky, Daniel; Bjørnestad, Victoria Ariel; Hodnik, Vesna; Anderluh, Gregor; Krengel, Ute

    2016-01-01

    Cholera is the prime example of blood-group-dependent diseases, with individuals of blood group O experiencing the most severe symptoms. The cholera toxin is the main suspect to cause this relationship. We report the high-resolution crystal structures (1.1–1.6 Å) of the native cholera toxin B-pentamer for both classical and El Tor biotypes, in complexes with relevant blood group determinants and a fragment of its primary receptor, the GM1 ganglioside. The blood group A determinant binds in the opposite orientation compared to previously published structures of the cholera toxin, whereas the blood group H determinant, characteristic of blood group O, binds in both orientations. H-determinants bind with higher affinity than A-determinants, as shown by surface plasmon resonance. Together, these findings suggest why blood group O is a risk factor for severe cholera. PMID:27082955

  9. Beyond the Sum of Parts: Voting with Groups of Dependent Entities.

    Science.gov (United States)

    Yarlagadda, Pradeep; Ommer, Björn

    2015-06-01

    The high complexity of multi-scale, category-level object detection in cluttered scenes is efficiently handled by Hough voting methods. However, the main shortcoming of the approach is that mutually dependent local observations are independently casting their votes for intrinsically global object properties such as object scale. Object hypotheses are then assumed to be a mere sum of their part votes. Popular representation schemes are, however, based on a dense sampling of semi-local image features, which are consequently mutually dependent. We take advantage of part dependencies and incorporate them into probabilistic Hough voting by deriving an objective function that connects three intimately related problems: i) grouping mutually dependent parts, ii) solving the correspondence problem conjointly for dependent parts, and iii) finding concerted object hypotheses using extended groups rather than based on local observations alone. Early commitments are avoided by not restricting parts to only a single vote for a locally best correspondence and we learn a weighting of parts during training to reflect their differing relevance for an object. Experiments successfully demonstrate the benefit of incorporating part dependencies through grouping into Hough voting. The joint optimization of groupings, correspondences, and votes not only improves the detection accuracy over standard Hough voting and a sliding window baseline, but it also reduces the computational complexity by significantly decreasing the number of candidate hypotheses. PMID:26357338

  10. Fission product margin in burnup credit analyses

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W. [OECD Halden Reactor Project (Norway)

    1996-03-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  14. Study of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    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)

  15. Compressive creep of simulated burnup fuel

    International Nuclear Information System (INIS)

    In order to study the nitride fuel mechanical properties, we measured the compressive steady state creep rates of uranium mononitride (UN) and UN containing neodymium which was simulated burnup fuel. The stress exponent n'' and the apparent activation energy ''Q'' of the creep rate were determined in the range of 27.5 ≤ σ ≤ 200.0 MPa and 950 ≤ T ≤ 1500 degC. (author)

  16. DELIGHT-6: one dimensional lattice burn-up code for high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    The code, DELIGHT-6, performs multi-group neutron spectrum calculation and provides few-group constans for succeeding core calculations. The main objective of the code is to serve as the lattice burn-up code for the core of a very high temperature gas-cooled reactor. The fuel rods of the reactor contain many coated fuel particles resulting double heterogeneous arrangement. The main calculational schema of DELIGHT-6 code is as follows; (1) Energy range for fast neutrons covers from 10 MeV to 2.38 eV and is divided into 61 fine groups. The thermal neutrons covers the rest of the energy range from 2.38 eV to 0 eV. Thermal spectrum is calculated by P1 or P0 approximation with 50 fine groups. (2) To treat resonance absorption, IR method is employed. (3) Zero and one dimensional models are available for the fuel lattice geometry and used for criticality and burn-up calculations. Collision probability method is adopted for the calculation of one dimensional model. (4) Shielding factor of burnable poison is calculated by collision probability method. (5) Other functions of the code are; 1. Spatial shielding factor calculation of 240Pu, 2. Calculation of neutron streaming effect caused by a gap or a hole in the fuel lattice, 3. Calculation of neutron flux distribution in the fuel lattice by diffusion theory, 4. Calculation of Xe and Sm absorption cross sections with burn-up. (6) Cross section library in both fast and thermal energy range is compiled from ENDF/B-4 except burn-up data of Xm, Sm and pseudo FPs which are supplied by ENDF/B-3. (7) The code provides the macroscopic group constants of fuel lattice with burn-up in CITATION input format. (jin)

  17. 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.

  18. OECD/NEA Burnup Credit Criticality Benchmark

    International Nuclear Information System (INIS)

    The report describes the final result of the phase-1A of the Burnup Credit Criticality Benchmark conducted by OECD/NEA. The phase-1A benchmark problem is an infinite array of a simple PWR spent fuel rod. The analysis has been performed for the PWR spent fuels of 30 and 40 GWd/t after 1 and 5 years of cooling time. In total, 25 results from 19 institutes of 11 countries have been submitted. For the nuclides in spent fuel, 7 major actinides and 15 major fission products (FP) are selected for the benchmark calculation. In the case of 30 GWd/t burnup, it is found that the major actinides and the major FPs contribute more than 50% and 30% of the total reactivity loss due to burnup, respectively. Therefore, more than 80% of the reactivity loss can be covered by 22 nuclides. However, the larger deviation among the reactivity losses by participants has been found for cases including EPs than the cases with only actinides, indicating the existence of relatively large uncertainties in FP cross sections. The large deviation seen also in the case of the fresh fuel has been found to reduce sufficiently by replacing the cross section library from ENDF-B/IV with that from ENDF-B/V and taking the known bias of MONK6 into account. (author)

  19. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, V.; Smirnov, A. [RRC Research Institute of Atomic Reactors, Dimitrovqrad (Russian Federation)

    1996-03-01

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

  20. 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.

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

    International Nuclear Information System (INIS)

    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

  2. Fuel performance at high burnup for water reactors

    International Nuclear Information System (INIS)

    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

  3. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations

    International Nuclear Information System (INIS)

    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 keff (''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. Burnup effects of MOX fuel pincells in PWR - OECD/NEA burnup credit benchmark analysis -

    International Nuclear Information System (INIS)

    The burnup effects were analyzed for various cases of MOX fuel pincells of fresh and irradiated fuels by using the HELIOS, MCNP-4/B, CRX and CDP computer codes. The investigated parameters were burnup, cooling time and combinations of nuclides in the fuel region. The fuel compositions for each case were provided by BNFL (British Nuclear Fuel Limited) as a part of the problem specification so that the results could be focused on the calculation of the neutron multiplication factor. The results of the analysis show that the largest saving effect of the neutron multiplication factor due to burnup credit is 30 %. This is mainly due to the consideration of actinides and fission products in the criticality analysis

  5. Group Interventions with Low-Income African American Women Recovering from Chemical Dependency.

    Science.gov (United States)

    Washington, Olivia G. M.; Moxley, David P.

    2003-01-01

    Presents finding from an investigation of two group therapy modalities involving 93 women with dependent children and limited education and income levels. An overview of intervention activities that participants found beneficial is presented. Programs were found to help participants develop a sense of community, reduce stress, improve…

  6. Prize Reinforcement Contingency Management for Cocaine Dependence: Integration with Group Therapy in a Methadone Clinic

    Science.gov (United States)

    Petry, Nancy M.; Martin, Bonnie; Simcic, Francis

    2005-01-01

    In this study, the authors evaluated a low-cost contingency management (CM) procedure for reducing cocaine use and enhancing group therapy attendance in 77 cocaine-dependent methadone patients. Patients were randomly assigned to 12 weeks of standard treatment or standard treatment with CM, in which patients earned the opportunity to win prizes…

  7. Efficacy of Group Motivational Interviewing (GMI) for Psychiatric Inpatients with Chemical Dependence

    Science.gov (United States)

    Santa Ana, Elizabeth J.; Wulfert, Edelgard; Nietert, Paul J.

    2007-01-01

    Dually diagnosed patients with chemical dependency and a comorbid psychiatric disorder typically show poor compliance with aftercare treatment, which may result in costly and pervasive individual and societal problems. In this study, the authors investigated the effect of adding motivational interviewing in a group format to standard treatment for…

  8. Symmetry group of point transformations for the time-dependent Schroedinger equation: Harmonic interactions among nucleons

    International Nuclear Information System (INIS)

    We have used Lie's method of extended group to obtain explicit forms of the generators and the structure of the maximal symmetry group of point transformations of the time-dependent Schroedinger equation for motions of nucleons interacting with two-body harmonic potential. The generators of the symmetry group correspond to different states of motion of the system. The maximal symmetry group is found to be a semidirect product of an infinite parameter Abelian invariant subgroup and a proper subgroup. For Z protons and N neutrons, this proper subgroup is a Lie group with 1/2[9Z(Z-1)+9N(N-1)+40] generators. Different nuclear modes of excitations have been assigned to the different generators. In particular the giant resonance mode and other collective modes of motion are shown to be consequences of the symmetry of the system

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

    International Nuclear Information System (INIS)

    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

  10. Implementation of burnup credit in spent fuel management systems

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  12. The US department of energy's transportation burnup credit program

    International Nuclear Information System (INIS)

    Aspects of the U. S. Department of Energy's (DOE's) transportation burnup credit program, the Department's motivation for conducting the program, and the status of burnup credit activities are presented. The benefits, technical, and regulatory considerations associated with using burnup credit for transport of irradiated nuclear fuel are discussed. The methods used in the DOE's actinide-only topical report are described in terms of the technical and regulatory issues. (authors)

  13. Multi-group formulation of the temperature-dependent resonance scattering model and its impact on reactor core parameters

    International Nuclear Information System (INIS)

    Highlights: • Multi-group formulation for exact neutron elastic scattering kernel is developed. • Up-scattering effects are incorporated in the cross-section data for heavy nuclei. • Effects on Doppler Temperature Coefficient (DTC) are demonstrated using DRAGON. • Results show an increase in DTC values by almost 10% for UOX and MOX LWR fuels. - Abstract: A multi-group formulation for the exact neutron elastic scattering kernel is developed. It incorporates the neutron up-scattering effects stemming from lattice atoms thermal motion and it accounts for them within the resulting effective nuclear cross-section data. The effects pertain essentially to resonant scattering off of heavy nuclei. The formulation, implemented into a standalone code, produces effective nuclear scattering data that are then supplied directly into the DRAGON lattice physics code where the effects on Doppler reactivity and neutron flux are demonstrated. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. The results show an increase in values of Doppler temperature feedback coefficients up to −10% for UOX and MOX LWR fuels compared to the corresponding values derived using the traditional asymptotic elastic scattering kernel. This paper also summarizes research performed to date on this topic

  14. Core burnup characteristics of high conversion light water reactor, (1)

    International Nuclear Information System (INIS)

    In order to evaluate core burnup characteristics of a high conversion light water reactor (HCLWR) with tight pitched lattice, core burnup calculation was made using two dimensional diffusion method. The volume ratio of moderator to fuel is about 0.8 in the reactor (HCLWR-J1) under study. The burnup calculations were carried out under the assumption of three batch and out-in fuel loading from the first cycle to the equilibrium cycle. A detailed evaluation was made for discharge burnup, conversion ratio, power distribution, and reactivity coefficients and so on. (author)

  15. Kelvin-probe force microscopy of the pH-dependent charge of functional groups

    Science.gov (United States)

    Stone, Alexander D. D.; Mesquida, Patrick

    2016-06-01

    Kelvin-probe Force Microscopy (KFM) is an established method to map surface potentials or surface charges at high, spatial resolution. However, KFM does not work in water, which restricts its applicability considerably, especially when considering common, functional chemical groups in biophysics such as amine or carboxy groups, whose charge depends on pH. Here, we demonstrate that the KFM signal of such groups taken in air after exposure to water correlates qualitatively with their expected charge in water for a wide range of pH values. The correlation was tested with microcontact-printed thiols exposing amine and carboxy groups. Furthermore, it was shown that collagen fibrils, as an example of a biological material, exhibit a particular, pH-sensitive surface charge pattern, which could be caused by the particular arrangement of ionizable residues on the collagen fibril surface.

  16. 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)

    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. 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. PMID:26509282

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  20. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  2. 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.

  3. Burnup calculation capability in the PSG2 / Serpent Monte Carlo reactor physics code

    International Nuclear Information System (INIS)

    The PSG continuous-energy Monte Carlo reactor physics code has been developed at VTT Technical Research Centre of Finland since 2004. The code is mainly intended for group constant generation for coupled reactor simulator calculations and other tasks traditionally handled using deterministic lattices physics codes. The name was recently changed from acronym PSG to 'Serpent', and the capabilities have been extended by implementing built-in burnup calculation routines that enable the code to be used for fuel cycle studies and the modelling of irradiated fuels. This paper presents the methodology used for burnup calculation. Serpent has two fundamentally different options for solving the Bateman depletion equations: 1) the Transmutation Trajectory Analysis method (TTA), based on the analytical solution of linearized depletion chains and 2) the Chebyshev Rational Approximation Method (CRAM), an advanced matrix exponential solution developed at VTT. The first validation results are compared to deterministic CASMO-4E calculations. It is also shown that the overall running time in Monte Carlo burnup calculation can be significantly reduced using specialized calculation techniques, and that the continuous-energy Monte Carlo method is becoming a viable alternative to deterministic assembly burnup codes. (authors)

  4. Fuel burnup calculation for HEU and LEU cores of Ghana MNSR

    International Nuclear Information System (INIS)

    Fuel burnup calculations have been performed using a computer program developed as part of this research work for both Highly Enriched Uranium (90.2 % U-235) and Low Enriched Uranium (12.6 % U-235) cores for Ghana Research Reactor-1 (GHARR-1). Fuel depletion analyses of the GHARR-1 core was also performed which provided an inventory of the actinides formed as a result of burnup. The effect of the production of plutonium isotopes with burnup on reactor operation was also estimated. A FORTRAN 95 code was written based on the three group model approach namely fast, resonance and slow (thermal) neutron reactions. The time rate of change of each fuel isotope density is given by a first order differential equation. A general solution for each fuel isotope rate equation was used as input for the computer code. These results are particularized to the case of constant power during a short time interval, during which the slow (thermal) neutron flux is considered constant. The results obtained for the HEU were in good agreement with those found in literature. Therefore, this code can be used to estimate the burnup of LEU fuel for core conversion from HEU to LEU. (au)

  5. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    International Nuclear Information System (INIS)

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    Highlights: • Increasing the time of nuclear fuel into reactor generates high burnup structure. • We analyze model to simulate high burnup scenarios for UO2 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

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

    International Nuclear Information System (INIS)

    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 149Sm, 151Sm, and 155Gd

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

    International Nuclear Information System (INIS)

    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

  10. Continuation of the VVER burnup credit benchmark. Evaluation of CB1 results, overview of CB2 results to date, and specification of CB3

    International Nuclear Information System (INIS)

    A calculational benchmark focused on VVER-440 burnup credit, similar to that of the OECD/NEA/NSC Burnup Credit Benchmark Working Group, was proposed on the 96'AER Symposium. Its first part, CB1, was specified there whereas the second part, CB2, was specified a year later, on 97'AER Symposium in Zittau. A final statistical evaluation is presented of CB1 results and summarizes the CB2 results obtained to date. Further, the effect of an axial burnup profile of VVER-440 spent fuel on criticality ('end effect') is proposed to be studied in the CB3 benchmark problem of an infinite array of VVER-440 spent fuel rods. (author)

  11. Point reactivity burnup code DELIGHT-4 for high temperature, gas-cooled reactor cells

    International Nuclear Information System (INIS)

    The code DELIGHT-4 has been developed for analizing burnup characteristics of the graphite moderated reactor cells and producing the few-group constants. Calculation models for the code are as follows: (1) The number of neutron energy groups is 61 for fast neutrons (10 MeV -- 2.38 eV) and 50 for thermal neutrons (2.38 eV -- 0 eV). (2) The doubly space-heterogeneous effect of fuel (dispersion of coated fuel particles in fuel compacts and regular array of fuel rods in graphite blocks) is considered in the calculation of resonance absorption. (3) The double heterogenity of burnable poison (dispersion of absorber grains in rods) can be considered. (4) The chemical binding effect of graphite is introduced in the scattering of thermal neutrons. (5) The calculations of criticality and burnup are by a few-energy-group models (up to 10 groups for both fast and thermal neutrons), and nuclide chains of thorium-uranium and uranium-plutonium are used for burnup calculation. (6) Neutron streaming effect through holes and gaps in cells can be considered in criticality calculation. (7) The flux distribution in cells can be calculated. The cell-averaged few group constants can be produced in card form for 1-D transport approximation code SLALOM, 2-D S sub( n) code TWOTRAN, 1-D diffusion code BRIQUET, 2-D diffusion code ZADOC-3 and 3-D diffusion code CITATION-DEGA. (author)

  12. Revised SWAT. The integrated burnup calculation code system

    International Nuclear Information System (INIS)

    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)

  13. 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)

  14. Effects of high burnup on spent-fuel casks

    International Nuclear Information System (INIS)

    Utility fuel managers have become very interested in higher burnup fuels as a means to reduce the impact of refueling outages. High-burnup fuels have significant effects on spent-fuel storage or transportation casks because additional heat rejection and shielding capabilities are required. Some existing transportation casks have useful margins that allow shipment of high-burnup fuel, especially the NLI-1/2 truck cask, which has been relicensed to carry pressurized water reactor (PWR) fuel with 56,000 MWd/ton U burnup at 450 days of cooling time. New cask designs should consider the effects of high burnup for future use, even though it is not commercially desirable to include currently unneeded capability. In conclusion, the increased heat and gamma radiation of high-burnup fuels can be accommodated by additional cooling time, but the increased neutron radiation source cannot be accommodated unless the balance of neutron and gamma contributions to the overall dose rate is properly chosen in the initial cask design. Criticality control of high-burnup fuels is possible with heavily poisoned baskets, but burnup credit in licensing is a much more direct means of demonstrating criticality safety

  15. Implementation of burnup credit in PWR spent fuel storage pools

    International Nuclear Information System (INIS)

    Implementation of burnup credit in spent fuel storage of LWR fuel at nuclear power plants is approved in Germany since the beginning of 2000. The burnup credit methods applied have to comply with the newly developed German criticality safety standard DIN 25471 passed in November 1999 and published in September 2000, cp. (orig.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2001-09-28

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

  19. In-reactor thermo-mechanical measurements on LWR fuel rods in the high burnup range

    International Nuclear Information System (INIS)

    The extension of fuel burn-up beyond previously accepted levels is currently being applied in varying degrees throughout the nuclear industry, with the aim of improving fuel economics and reducing the spent fuel volume. So it is necessary that the current fuel knowledge base should be extended. Modifications of fuel rod/assembly concepts, together with fuel management schemes, should be gradually implemented so that the operation of power reactors becomes even more reliable and flexible than it is today. Extrapolation to extended burn-up levels does not cause concern but will have to be made in steps, in order to demonstrate expected performance trends. The fuel testing programmes at the OECD Halden Reactor Project have over the years significantly contributed to the understanding of LWR fuel behaviour in the high burn-up range. A broad range of versatile and integrated in-reactor test rigs and high pressure loops have been developed which allow simulations of LWR irradiation conditions, comparative testing of alternative fuel rod designs and use of test segments pre-irradiated in power reactors. A number of in-core instruments and experimental techniques have been developed for detailed investigations of various aspects related to the thermal behaviour, fission product release and mechanical response of high burn-up LWR fuel rods, under a variety of operating conditions. The paper reviews recent measurements in the area of burnup-dependent steady-state and transient thermal behaviour of fuel rods, intermixing of fission and helium filler gases in the pellet cladding gap, fission gas release kinetics under changing heat loads and power excursions (burst release) and dimensional changes of fuel rods subjected to cyclic load changes. (author). 14 refs, 12 figs

  20. Finnish contribution to the CB4 burnup credit benchmark

    International Nuclear Information System (INIS)

    The CB4 phase of the WWER burnup credit benchmark series studies the effect of flat and realistic axial burnup profiles on the multiplication factor of a conceptual WWER cask loaded with spent fuel. The benchmark was calculated at VTT Energy with MCNP4C, using mainly ENDF/B-V1 cross sections. According to the calculation results the effect of the axial homogenization on the keff estimate is complex. At low burnups the use of a axial profile overestimates keff but at high burnups the reverse is the case. Ignoring fission products leads to conservative keff and the effect of axial homogenization on the multiplication factor is similar to a reduction of the burnup (Authors)

  1. Automated generation of burnup chain for reactor analysis applications

    International Nuclear Information System (INIS)

    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)

  2. Probabilistic assessment of dry transport with burnup credit

    International Nuclear Information System (INIS)

    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)

  3. Thermonuclear burn-up in deuterated methane $CD_4$

    CERN Document Server

    Frolov, Alexei M

    2010-01-01

    The thermonuclear burn-up of highly compressed deuterated methane CD$_4$ is considered in the spherical geometry. The minimal required values of the burn-up parameter $x = \\rho_0 \\cdot r_f$ are determined for various temperatures $T$ and densities $\\rho_0$. It is shown that thermonuclear burn-up in $CD_4$ becomes possible in practice if its initial density $\\rho_0$ exceeds $\\approx 5 \\cdot 10^3$ $g \\cdot cm^{-3}$. Burn-up in CD$_2$T$_2$ methane requires significantly ($\\approx$ 100 times) lower compressions. The developed approach can be used in order to compute the critical burn-up parameters in an arbitrary deuterium containing fuel.

  4. Investigation of burnup credit implementation for BWR fuel

    International Nuclear Information System (INIS)

    Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumptions made to define the axial profile of the burnup and void fraction (for BWR), to determine the composition of the irradiated fuel and to compute the criticality simulation. In the framework of Burnup Credit implementation for BWR fuel, this paper proposes to investigate part of these items. The studies presented in this paper concern: the influence of the burnup and of the void fraction on BWR spent fuel content and on the effective multiplication factor of an infinite array of BWR assemblies. A code-to-code comparison for BWR fuel depletion calculations relevant to Burnup Credit is also performed. (authors)

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

    International Nuclear Information System (INIS)

    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 UO2 critical experiments to determine the criticality safety limits on the neutron multiplication factor, keff. 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)

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

    International Nuclear Information System (INIS)

    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

  7. A questionnaire on survival of kittens depending on the blood groups of the parents.

    Science.gov (United States)

    Axnér, Eva

    2014-10-01

    Cats more than 2 months of age have alloantibodies against the blood type antigen that they do not possess. Maternal antibodies, including alloantibodies against blood groups, are transferred to the kittens' systemic circulation when they suckle colostrum during the first 12-16 h after birth. If kittens with blood group A or AB nurse from a mother with blood group B they may develop neonatal isoerythrolysis (NI). Breeders can prevent kittens at risk of NI from nursing during the first 16-24 h; after this period it is safe to let them nurse. Kittens depend, however, on the passive transfer of antibodies from the colostrum for early protection against infections. Although it is known that kittens deprived of colostrum will also be deprived of passive systemic immunity, it is not known if this will affect their health. Therefore, the aim of this study was to evaluate kitten mortality in litters with B-mothers and A-fathers compared to litters with A-mothers. In addition, the aim was to evaluate the effects of colostrum deprivation on the health of the mothers, and the breeders' opinions and experiences of these combinations of breedings. A web-based questionnaire was constructed and distributed to breeders. The results indicate that there is no difference in mortality between planned litters that have mothers with blood group A and litters with mothers that have blood group B and fathers that have blood group A. When managing blood group incompatibility in cat all factors affecting the health of the cats, including genetic variation, should be considered. PMID:24423812

  8. Use of burnup credit in criticality safety design analysis of spent fuel storage systems

    International Nuclear Information System (INIS)

    Full text: It is well known that the use of Burnup Credit (BUC) in criticality safety design analysis of spent fuel storage systems significantly impacts the design of the system. BUC is defined as the consideration of the change in the fuel's isotopic composition and hence in its reactivity due to the irradiation of the fuel. Using BUC means to identify that isotopic composition and hence that burnup which just results in the maximum neutron multiplication factor allowable for the system, including all mechanical and calculational uncertainties. This burnup is the minimum burnup necessary for fuel to be loaded in the system. Since the isotopic composition at given burnup depends on the initial enrichment of the fuel, the minimum burnup is usually given as a function of the initial enrichment. The graph of this function is commonly named as 'loading curve'. Thus, application of BUC to a spent fuel storage system consists in implementation of three key steps: Determination of the isotopic composition as a function of burnup and initial enrichment; Criticality calculation and evaluation of the loading curve; Quantification and verification of the actual burnup of the fuel to be loaded into the system. The main considerations of the first and the second step will be discussed. The isotopic composition is predicted by means of depletion calculations. To perform such calculations the parameters describing the fuel design characteristics and the fuel depletion conditions have to be defined. In addition the cooling time that may be credited (e.g., in BUC applications to spent fuel storage/transport cask systems) has to be specified. These parameters will be discussed with particular attention being given to the sensitivity of the neutron multiplication factor of the storage system to variations in the parameters and conditions characterizing the depletion conditions. These parameters and conditions are: Specific power and operating history, fuel temperature, moderator

  9. PLD-IDMS studies towards direct measurement of burn-up of nuclear fuel

    International Nuclear Information System (INIS)

    A method based on Pulsed laser deposition followed by Isotope dilution mass spectrometric method is evaluated towards the possibility of direct measurement of burn up of nuclear fuel and also to find out spatial distribution of burn-up along the pellet. The wave length dependent results show larger error with 1064 nm, compared to 532 nm laser beam. Much less error is expected with shorter wave length and shorter pulse width laser beam. Further work is being carried out in this direction

  10. DELIGHT-6(revised): one dimensional lattice burnup code for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    The code, DELIGHT-6, performs the multi-group neutron spectrum calculation and provides the few-group constants for burnup calculations of a high temperature gas-cooled reactor core, whose fuel elements containing many coated fuel particles are arranged in double heterogeneity. The main revisions in the DELIGHT-6 (Revised) are as follows; (1)The option of a sphere fuel cell calculation is added for the core design of pebble bed type high temperature gas-cooled reactor. (2)The yield and decay constants of fission products for burnup calculation is revised. (3)The following auxiliary functions are added; (i) Automatic calculation of averaged atom number density in the fuel region, (ii) Estimation of local neutron flux distribution (disadvantage factor), (iii) Preparation of the data for the fine mesh core calculation. (author)

  11. Salt effects on lamellar repeat distance depending on head groups of neutrally charged lipids.

    Science.gov (United States)

    Hishida, Mafumi; Yamamura, Yasuhisa; Saito, Kazuya

    2014-09-01

    Change in lamellar repeat distances of neutrally charged lipids upon addition of monovalent salts was measured with small-angle X-ray scattering for combinations of two lipids (PC and PE lipids) and six salts. Large dependence on lipid head group is observed in addition to those on added cation and anion. The ion and lipid dependences have little correlation with measured surface potentials of lipid membranes. These results indicate that the lamellar swelling by salt is not explained through balance among interactions considered previously (van der Waals interaction, electrostatic repulsion emerged by ion binding, etc.). It is suggested that effect of water structure, which is affected by not only ions but also lipid itself, should be taken into account for understanding membrane-membrane interactions, as in the Hofmeister effect. PMID:25126900

  12. A closely related group of RNA-dependent RNA polymerases from double-stranded RNA viruses.

    OpenAIRE

    Bruenn, J A

    1993-01-01

    Probably one of the first proteinaceous enzymes was an RNA-dependent RNA polymerase (RDRP). Although there are several conserved motifs present in the RDRPs of most positive and double-stranded RNA (dsRNA) viruses, the RDRPs of the dsRNA viruses show no detectable sequence similarity outside the conserved motifs. There is now, however, a group of dsRNA viruses of lower eucaryotes whose RDRPs are detectably similar. The origin of this sequence similarity appears to be common descent from one o...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-01

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

  14. 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. PMID:26582023

  15. Do the stellar populations of the brightest two group galaxies depend on the magnitude gap?

    CERN Document Server

    Trevisan, M; Khosroshahi, H G

    2016-01-01

    We investigate how the stellar populations of first and second brightest group galaxies (respectively BGGs and SBGGs) vary as a function of the magnitude gap, {\\Delta}M_12, using an SDSS-based sample of 569 groups with elliptical BGGs. The sample is complete in redshift, luminosity and for {\\Delta}M_12 up to 2.5 mag, and contains 75 optical fossil groups (FGs, with {\\Delta}M_12 > 2.0 mag). We determine ages, metallicities, and star formation histories (SFHs) of BGGs and SBGGs using the STARLIGHT code with two single stellar population (SSP) models, one of which (MILES) leads to significantly more extended SFHs than the other (BC03). After removing the dependence with stellar mass, there is no correlation with magnitude gap of BGG ages, metallicities, and SFHs derived with the BC03 model. However, with the MILES model, the BGGs in FGs appear to have more extended SFHs than those in regular groups. But this signature with MILES is not seen in the colours, specific star formation rates nor in the 4000 A breaks, ...

  16. The commercial impact of burnup increase

    International Nuclear Information System (INIS)

    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

  17. High burnup models in computer code fair

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  19. Analysis of burnup credit on spent fuel storage

    International Nuclear Information System (INIS)

    Chemical analyses were carried out on high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins. Measured data of the composition of nuclides from 234U to 242Pu were used for evaluation of ORIGEN-2/82 code. Criticality calculations were executed for the casks which were being designed to store 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for (1) axial and horizontal profiles of burnup, and void history (BWR), (2) operational histories such as control rod insertion history, BPR insertion history and others, and (3) calculational accuracy of ORIGEN-2/82 code on the composition of nuclides. Present evaluation shows that introduction of burnup credit has a substantial merit in criticality safety analysis of the cask, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for present reactivity bias evaluation and showed a possibility of simplifying the reactivity bias evaluation in burnup credit. Finally, adapting procedures of burnup credit such as the burnup meter were evaluated. (author)

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

    International Nuclear Information System (INIS)

    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)

  1. Current Status of Burnup Evaluation for Test Fuel at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Park, Seung Jae; Shin, Yoon Taeg; Choo, Kee Nam; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For the research reactor, 8 mini plate fuels were irradiation-tested during 4 irradiation cycles. 2 more irradiation capsules were fabricated for additional test of plate type fuel. Also fission Mo target for the performance verification and the demonstration of Mo-99 extraction process will be irradiated at HANARO. It is important to evaluate the burnup history of test fuel. The burnup of test fuel has been calculated using HANARO Fuel Management System (HANAFMS). Although it is proper to evaluate the burnup of HANARO fuel, it is difficult to accurately calculate the burnup of test fuel due to the limitation of HANAFMS model. Therefore, the improvement of burnup evaluation for the recent irradiated test fuel is conducted and reported in this paper. To evaluate the burnup of test fuel, HANAFMS has been used; however, HANAFMS model is not proper to apply plate type fuel. Therefore, MCNP burned core model was developed for HAMP-1 burnup calculation. Throughout the comparison of fuel assembly power, MCNP burned core model showed the good agreement with HANAFMS.

  2. Method for calculation of global sensitivity indices for few-group cross-section-dependent problems

    International Nuclear Information System (INIS)

    The variance based global sensitivity analysis technique is robust, has a wide range of applicability and provides accurate sensitivity information for most models. However, it requires input variables to be mutually independent. A modification to this technique that allows one to deal with input variables that are block-wise correlated and normally distributed is presented. The focus of this study is the application of the modified global sensitivity analysis technique to calculations of reactor parameters that are dependent on multigroup or few-group neutron cross-sections. The result of the sensitivity analysis is obtained in terms of the global sensitivity indices, which can be used for characterising the contribution of uncertainties from the input cross-sections or their groups to the uncertainty of the calculated reactor parameter. The main effort in this work, besides presenting the theoretical background, is in establishing a method for a practical numerical calculation of the global sensitivity indices. The implementation of the method involves the calculation of multi-dimensional integrals, which can be prohibitively expensive to compute. Numerical techniques specifically suited to the evaluation of multidimensional integrals namely Monte Carlo and sparse grids methods are used, and their efficiency is compared. The method is illustrated and tested on a two-group cross-section dependent problem. In all the cases considered the results obtained with sparse grids achieved much better accuracy while using a significantly smaller number of samples. This aspect is addressed in a mini-study and a preliminary explanation of the results obtained is given. (author)

  3. Fuel burnup characteristics for the NRU research reactor

    International Nuclear Information System (INIS)

    The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U3Si, consisting of particles of U3Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)

  4. Fuel burnup characteristics for the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C., E-mail: leungt@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    The driver fuel of the NRU research reactor at AECL, Chalk River is a low enriched uranium (LEU) fuel alloy of Al-61 wt% U{sub 3}Si, consisting of particles of U{sub 3}Si dispersed in a continuous aluminum matrix, with 19.8% U235 in uranium. This paper describes the burnup characteristics for this type of fuel in NRU, including the determination of fuel depletion using the neutronic simulation code TRIAD, comparisons between simulated and measured burnup values, and the regulatory licensing operational average fuel burnup limit. (author)

  5. Calculation study of TNPS spent fuel pool using burnup credit

    International Nuclear Information System (INIS)

    Exampled by the spent fuel pool of TNPS which is consist of 2 × 5 fuel storage racks, the spent fuel high-density storage based on burnup credit (BUC) and related criticality safety issues were studied. The MONK9A code was used to analyze keff, of different enrichment fuels at different burnups. A reference loading curve was proposed in accordance with the system keff's changing with the burnup of different initially enriched nuclear fuels. The capacity of the spent fuel pool increases by 31% compared with the one that does not consider BUC. (authors)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    It is clear that in order to maintain competitiveness with UO2 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 UO2 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)

  8. Burnup credit demands for spent fuel management in Ukraine

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Nazirzadeh, M.; Ghasemizad, A. [Department of Physics, University of Guilan, 41335-1914 Rasht (Iran, Islamic Republic of); Khanbabei, B. [School of Physics, Damghan University, 36716-41167 Damghan (Iran, Islamic Republic of)

    2015-12-15

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

  10. Burnup calculations of TR-2 Research Reactor with Monteburns Monte Carlo Code

    International Nuclear Information System (INIS)

    Full text: In this study, some neutronic calculations of first and second core cycles of 5 MW pool type TR-2 Research Reactor have been performed using Multi-Step Monte Carlo Burnup Code System MONTEBURNS and the results were compared with the values of experiments and other codes. Time dependent keff distribution and burnup ratios belong to first and second core cycles of TR-2 Research Reactor were compared and quite good consistence in the results were observed. After modeling the first and second core cycles of TR-2 with MCNP5 Monte Carlo code, MCNP5 used in MONTEBURNS code has been parallelized in 8 HP ProLiant BL680C G5 systems with 4 quad-core Intel Xeon E7330 CPU, utilizing the MPI parallel protocol and simulations were performed on the 128 cores Linux parallel computing machine system. The computation time was reduced by parallelization of MONTEBURNS which uses MCNP in many steps. (authors)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2015-12-01

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

  13. ABRAC: A microcomputer-based Fortran code for multi-cyle burnup

    International Nuclear Information System (INIS)

    Pressurized-water reactors have reactor physics and fuel management characteristics which are very amenable to simplified analysis. Given models which account for the dominant features of core and fuel performance, it is possible to rapidly perform relatively accurate scoping studies of many years of reactor operation in just a few hours on a modern (386-class) microcomputer. Models are described for burnup-dependent cross-section generation, for burnup of fuel under irradiation, and for computation of radial power distributions in hexagonal geometry assuming hexagonal fuel assemblies. Comparisons with more elaborate methods are given in order to validate the models, and to quantify the accuracy of the results. 16 refs., 5 figs., 5 tabs

  14. Cladding metallurgy and fracture behavior during reactivity-initiated accidents at high burnup

    International Nuclear Information System (INIS)

    High-burnup fuel failure during a reactivity-initiated accident has been the subject of safety-related concern. Because of wide variations in metallurgical and simulation test conditions, it has been difficult to understand the complex failure behavior from major tests in NSRR and CABRI reactors. In this paper, a failure model based on fracture toughness and microstructural characteristics is proposed in which fracture toughness of high-burnup cladding is assumed to be sensitive to temperature and exhibit ductile-brittle transition phenomena similar to those of irradiated bcc alloys. Significant effects of temperature and shape of the pulse are predicted when a simulated test is conducted near the material's transition temperature. Temperature dependence of fracture toughness is, in turn, sensitive to cladding microstructure such as density, distribution, and orientation of hydrides, oxygen distribution in the metallic phase, and irradiation-induced damage. Because all these factors are strongly influenced by corrosion, the key parameters that influence susceptibility to failure are oxide layer thickness and hydriding behavior. Therefore, fuel failure is predicted to be strongly dependent on cladding axial location as well as on burnup. 10 figs, 21 refs

  15. Generalization of the time-dependent numerical renormalization group method to finite temperatures and general pulses

    Science.gov (United States)

    Nghiem, H. T. M.; Costi, T. A.

    2014-02-01

    The time-dependent numerical renormalization group (TDNRG) method [Anders et al., Phys. Rev. Lett. 95, 196801 (2005), 10.1103/PhysRevLett.95.196801] offers the prospect of investigating in a nonperturbative manner the time dependence of local observables of interacting quantum impurity models at all time scales following a quantum quench. Here, we present a generalization of this method to arbitrary finite temperature by making use of the full density matrix approach [Weichselbaum et al., Phys. Rev. Lett. 99, 076402 (2007), 10.1103/PhysRevLett.99.076402]. We show that all terms in the projected full density matrix ρi →f=ρ+++ρ--+ρ+-+ρ-+ appearing in the time evolution of a local observable may be evaluated in closed form at finite temperature, with ρ+-=ρ-+=0. The expression for ρ-- is shown to be finite at finite temperature, becoming negligible only in the limit of vanishing temperatures. We prove that this approach recovers the short-time limit for the expectation value of a local observable exactly at arbitrary temperatures. In contrast, the corresponding long-time limit is recovered exactly only for a continuous bath, i.e., when the logarithmic discretization parameter Λ →1+. Since the numerical renormalization group approach breaks down in this limit, and calculations have to be carried out at Λ >1, the long-time behavior following an arbitrary quantum quench has a finite error, which poses an obstacle for the method, e.g., in its application to the scattering-states numerical renormalization group method for describing steady-state nonequilibrium transport through correlated impurities [Anders, Phys. Rev. Lett. 101, 066804 (2008), 10.1103/PhysRevLett.101.066804]. We suggest a way to overcome this problem by noting that the time dependence, in general, and the long-time limit, in particular, become increasingly more accurate on reducing the size of the quantum quench. This suggests an improved generalized TDNRG approach in which the system is time

  16. Development and validation of burnup function in reactor Monte Carlo RMC

    International Nuclear Information System (INIS)

    This paper presents the burnup calculation capability of RMC, which is a new Monte Carlo (MC) neutron transport code developed by Reactor Engineering Analysis Laboratory (REAL) in Tsinghua University of China. Unlike most of existing MC depletion codes which explicitly couple the depletion module, RMC incorporates ORIGEN 2.1 in an implicit way. Different burn step strategies, including middle-of-step approximation and predictor-corrector method, are adopted by RMC to assure accuracy under large step size. RMC employs a spectrum-based method of tallying one-group cross section, which can considerably save computational time with negligible accuracy loss. According to validation results of benchmarks and examples, it is proved that the burnup function of RMC performs quite well in accuracy and efficiency. (author)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

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

  19. A combined 1D/3D fuel burnup analysis of generation IV light water reactor IRIS

    International Nuclear Information System (INIS)

    A combined 1D/3D methodology for the fuel burnup analysis of generation IV light water reactors with thin boron coating that covers the fuel rods is described in this paper. This methodology is founded on three approximations. The first approximation assumes that the problem of fuel depletion in the entire 3D core can be resolved into two independent problems. One is a 3D Monte Carlo evolution of power distribution in large volumes (nodes) with the KENO-V.a code, and the other is a transport method evolution of burnup dependent fuel composition in 1D Wigner-Seitz cell for each node independently. With the second approximation, the time-dependent fuel composition in the node (e.g., in the fuel assembly) is calculated by using a 1D fuel depletion analysis with the SAS2H control module from the SCALE-4.4a code system. The third approximation involves smearing the boron coating with the clad (by volume homogenization). The proposed SAS2H/KENO-V.a methodology is verified for the case of 2D x-y model of IRIS 15x15 fuel assembly (with a reflective boundary condition) by using two well benchmarked code systems. The first one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. It has been found that the proposed SAS2H/KENO-V.a methodology gives a satisfactory accuracy for keff and nuclide composition. Finally, this methodology was applied for 3D burnup analysis of IRIS-1000 benchmark≠44 core. Detailed keff and power density evolution with burnup are reported. (author)

  20. A mechanism for the dependence of sunspot group tilt angles on cycle strength

    CERN Document Server

    Işık, Emre

    2015-01-01

    The average tilt angle of sunspot groups emerging throughout the solar cycle determines the net magnetic flux crossing the equator, which is correlated with the strength of the subsequent cycle. I suggest that a deep-seated, non-local process can account for the observed cycle-dependent changes in the average tilt angle. Motivated by helioseismic observations indicating cycle-scale variations in the sound speed near the base of the convection zone, I determined the effect of a thermally perturbed overshoot region on the stability of flux tubes and on the tilt angles of emerging flux loops. I found that 5-20 K of cooling is sufficient for emerging flux loops to reproduce the reported amplitude of cycle-averaged tilt angle variations, suggesting that it is a plausible effect responsible for the nonlinearity of the solar activity cycle.

  1. 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

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

    International Nuclear Information System (INIS)

    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

  3. Parametric neutronic analyses related to burnup credit cask design

    International Nuclear Information System (INIS)

    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 keff 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

  4. 2005 status and future of burnup credit in the USA

    International Nuclear Information System (INIS)

    At the beginning of 2005 in the USA burnup credit is licensed for PWR and BWR spent fuel pools, is under license review for a transport cask, is under discussion for disposal criticality. Two basic approaches exist for burnup credit. The first approach, which is licensed for spent fuel pools, utilizes criticality experience with spent fuel that has not been chemically assayed. The second approach to burnup credit comes from utilizing chemical assay data to validate the depletion calculations and then clean critical experiments to validate the criticality calculation. A burnup credit standard (ANS/ANSI-8.27) is under development where the two approaches are actively discussed. Issues related to the two approaches are presented as well as possible ways of resolving the issues. (author)

  5. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

    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)

  6. The comparison of recreative activities of 11-15 age group depending on different regions

    Directory of Open Access Journals (Sweden)

    Adem Pala

    2012-06-01

    Full Text Available The purpose of this study is to compare the recreative activities of 11-15 age group depending on different Regions. In this respect, 227 children from Southeastern part of Turkey and 262 children from Marmara region of Turkey participated in this study in a volunteered way.The questionnaire adapted from previous studies was implemented on these participants. Afterwards, the data were analyzed in terms of frequency, percentage and independent group of t-test statistics. Data were analyzed with the help of SPSS 15.0 and the significance rate was determined 0,05. Consequently, significantly different result were found (p0,05.According to the findings of this research, 37,9 % of the participant from Southeastern part of Turkey chose “sports” option of the item of “what they do in their leisure time”, 53,3 % of participant within same region chose “football” option of the item of “which recreative activities they regularly deal with”. 51,5 % of the participant from Marmara region chose “sports” option of the former question: Additionally , 25,6 % of item chose “football”, 21,5 % of them chose “badminton” and 12,8 % of them chose “basketball” options of the latter question.

  7. 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. PMID:25885332

  8. Rice rhizodeposition and its utilization by microbial groups depends on nitrogen fertilization

    Science.gov (United States)

    Ge, Tida; Zhu, Zhenke; Wu, Jinshui

    2016-04-01

    Rhizodeposited carbon (C) has received considerable attention because it plays an important role in regulating soil C sequestration and global C cycling, and represents the main C source for rhizosphere microorganisms. However, limited information exists on the utilization of rhizodeposited C by different microbial groups, its role in the turnover of soil organic matter (SOM) pools in rice paddies and how this is influenced by nitrogen (N) fertilization. Rice (Oryza sativa L.) was grown in soil at one of five N fertilization rates (0, 10, 20, 40, or 60 mg N kg‑1 soil) and then continuously labeled by exposure to a 13CO2 atmosphere for 18 days. The utilization of root-derived C by microbial groups within the rhizosphere was assessed by following the incorporation of 13C into phospholipid fatty acids (PLFAs). Rice shoot and root biomass strongly increased with N fertilization rate. Rhizodeposition was greater, but total 13C incorporation into microorganisms was lower, in N-fertilized soils than in unfertilized soil. The contribution of root-derived 13C to SOM formation increased with root biomass. The roots tended to grow into large aggregates (0.25-2.0 mm diameter), and N fertilization stimulated incorporation of 13C into these macroaggregates, presumably due to the relatively high root biomass. The ratio of 13C in soil pools (SOM, microbial biomass) to 13C in roots decreased as a result of N fertilization. N fertilization increased 13C incorporation into fungi (18:2ω6, 9c, 18:1ω9c), AM fungi (16:1ω5c), and actinomycetes (10Me 16:0, 10Me 18:0), but decreased 13C incorporation into Gram-positive (i14:0, i15:0, a15:0, i16:0, i17:0, a17:0) and Gram-negative (16:1ω7c, 18:1ω7c, cy17:0, cy19:0) bacteria. Thus, the uptake and processing of root-derived C by microbial groups depended on soil N status. Relative to the unfertilized controls, the contribution of rhizodeposited-C to SOM and microorganisms was increased by low to intermediate N fertilization rates, but

  9. EPRI R and D perspective on burnup credit

    International Nuclear Information System (INIS)

    'Burnup credit' refers to taking credit for the burnup of nuclear fuel in the performance of criticality safety analyses. Historically, criticality safety analyses for transport of spent nuclear fuel have assumed the fuel to be unirradiated (i.e. 'fresh' fuel). In 1999, the U.S. Nuclear Regulatory Commission (NRC) Spent Fuel Project Office issued Interim Staff Guidance - 8 (ISG-8) with recommendations for the use of burnup credit in storage and transportation of pressurized water reactor (PWR) spent fuel. The use of burnup credit offers an opportunity to reduce the number of spent nuclear fuel shipments by ∼30%. A simple analysis shows that the increased risk of a criticality event associated with properly using burnup credit is negligible. Comparing this negligible risk component with the reduction in common transport risks due to the reduced number of spent fuel shipments (higher capacity casks for transporting PWR spent fuel) leads to the conclusion that using 'burnup credit' is preferable to using the 'fresh fuel' assumption. A specific objective of the EPRI program is to support the Goals of the U.S. Industry. These goals are consistent with the original U.S. Department of Energy (DOE) goal defined in 1988: a burnup credit methodology that takes credit for the negative reactivity that is practical (all fissile actinides, most neutron absorbing actinides, and a subset of the fission products that account for the majority of the available credit from all fission products). The determination of the optimum number of fission products to consider in a practical burnup credit methodology validates the approach advocated by researchers from France to first focus on a handful of isotopes that include Sm-149; Rh-103; Nd-143; Gd-155; and Sm-152. (author)

  10. 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)

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

    International Nuclear Information System (INIS)

    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)

  12. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  14. Ductile-to-brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions

    Science.gov (United States)

    Billone, M. C.; Burtseva, T. A.; Einziger, R. E.

    2013-02-01

    Structural analyses of dry casks containing high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). A test procedure was developed to simulate the effects of drying-storage temperature histories. Following drying-storage simulation, samples were subjected to ring-compression test (RCT) loading, which was used as a ductility screening test and to simulate pinch-type loading that may occur during cask transport. RCT samples with 50% wall cracking were assessed as brittle. Prior to testing high-burnup cladding, many tests were conducted with pre-hydrided Zircaloy-4 (Zry-4) and ZIRLO™ to determine target 400 °C hoop stresses for high-burnup rodlets. Zry-4 cladding segments, from a 67-GWd/MTU fuel rod, with 520-620 wppm hydrogen and ZIRLO™ cladding segments from a 70-GWd/MTU fuel rod, with 350-650 wppm hydrogen were defueled and tested. Following drying-storage simulation, the extent of radial-hydride precipitation was characterized by the radial-hydride continuity factor. It was found that the DBTT was dependent on: cladding material, irradiation conditions, and drying-storage histories (stress at maximum temperature). High-burnup ZIRLO™ exhibited higher susceptible to radial-hydride formation and embrittlement than high-burnup Zry-4. It was also observed that uniformly pre-hydrided, non-irradiated cladding was not a good surrogate for high-burnup cladding because of the high density of circumferential hydrides across the wall and the high metal-matrix ductility for pre-hydrided cladding.

  15. Instrumentation for measuring the burnup of spent nuclear fuel

    International Nuclear Information System (INIS)

    Many different methods or procedures have been developed to measure reactivity of fissil materials. Few of these, however, have been designed specifically for light water reactor fuel or have actually been used to measure the reactivity of spent fuel. The methods that have been used to make measurements of related systems are the 252Cf source-driven noise analysis method, a noise analysis method using natural neutron sources, subcritical assembly measurements, and pulsed neutron techniques. Several different approaches to directly measuring burnup have been developed by various organizations. The experimental work on actual spent nuclear fuel utilizing reactivity measurement techniques is insufficient to provide conclusive evidence of the applicability of these techniques for verifying fuel burnup. The work with burnup meters indicates, however, that good correlations can be obtained with any of the systems. A burnup meter's primary function would be a secondary assurance that the administrative records are not grossly in error. Reactivity measurements provide information relating to the reactivity of the fuel only under the conditions measured. Criticality prevention design requirements will necessitate that casks accommodate a minimum burnup level for a given initial enrichment (i.e., a maximum reactivity). Direct measurement of the burnup will enable an easy determination of whether a particular fuel assembly can be shipped in a specific cask with a minimum number of additional correlations

  16. Burnup credit considerations in dry spent-fuel storage licensing

    International Nuclear Information System (INIS)

    Burnup credit has been allowed in reactor basin spent-fuel storage at pressurized water reactors for a number of years. However, such storage occurs under strict administrative, procedural, and design controls. In recent years, dry spent-fuel storage cask vendors have expressed interest in designing cask fuel baskets with allowance for burnup credit. At last year's American Nuclear Society Winter Meeting, an ad hoc session was organized and authorized on burnup credit for dry storage and transportation casks. It has become clear that some utilities are interested in burnup credit for dry storage designs. Given this, the US Nuclear Regulatory Commission (NRC) staff is examining the technical issues involved in allowing burnup credit. Analytical work focused on the development of branch technical positions for determination of burnup credit for dry spent-fuel storage technology designs has begun. Procedural and administrative issues will be examined, based on licensing experience, and will also be the subject of branch technical positions. At an appropriate time, preparation of regulatory guides will be considered

  17. Application of Candle burnup to small fast reactor

    International Nuclear Information System (INIS)

    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)

  18. Research on irradiation behavior of superhigh burnup fuel

    International Nuclear Information System (INIS)

    In Japan Atomic Energy Research Institute, the special team for LWR future technology development project was organized in Tokai Research Establishment from October, 1991 to the end of fiscal year 1993. Due to the delay of the introduction of fast reactors, LWRs are expected to be used for considerably long period also in 21st century, therefore, it aimed at the further advancement of LWRs, and as one of its embodiments, the concept of superhigh burnup fuel was investigated. The superhigh burnup fuel aims at the attainment of 100 GWd/t burnup, and it succeeded the achievement of the conceptual design study on 'superlong life LWRs'. It is generally recognized that the development of the new material that substitutes for zircaloy is indispensable for superhigh burnup fuel. The concept of superhigh burnup core and the specification of fuel, the research and development of superhigh burnup fuel, the research on the irradiation behavior and irradiation damage of fuel and the damage by ion irradiation, and the method and the results of the irradiation experiment using a tandem accelerator are reported. (K.I.)

  19. Research on irradiation behavior of superhigh burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1995-03-01

    In Japan Atomic Energy Research Institute, the special team for LWR future technology development project was organized in Tokai Research Establishment from October, 1991 to the end of fiscal year 1993. Due to the delay of the introduction of fast reactors, LWRs are expected to be used for considerably long period also in 21st century, therefore, it aimed at the further advancement of LWRs, and as one of its embodiments, the concept of superhigh burnup fuel was investigated. The superhigh burnup fuel aims at the attainment of 100 GWd/t burnup, and it succeeded the achievement of the conceptual design study on `superlong life LWRs`. It is generally recognized that the development of the new material that substitutes for zircaloy is indispensable for superhigh burnup fuel. The concept of superhigh burnup core and the specification of fuel, the research and development of superhigh burnup fuel, the research on the irradiation behavior and irradiation damage of fuel and the damage by ion irradiation, and the method and the results of the irradiation experiment using a tandem accelerator are reported. (K.I.).

  20. Triton burnup study in JT-60U

    International Nuclear Information System (INIS)

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

  1. Monte Carlo burnup analysis code development and application to an incore thermionic space nuclear power system

    International Nuclear Information System (INIS)

    In the design of the incore thermionic reactor system developed under the Advanced Thermionic Initiative (ATI), the fuel is highly enriched uranium dioxide and the moderating medium is zirconium hydride. The traditional burnup and fuel depletion analysis codes have been found to be inadequate for these calculations, largely because of the material and geometry modeled and because the neutron spectra assumed for the codes such as LEOPARD and ORIGEN do not even closely fit that for a small, thermal reactor using ZrH as moderator. More sophisticated codes such as the transport lattice type code WIMS often lack some materials, such as ZrH. Thus a new method which could accurately calculate the neutron spectrum and the appropriate reaction rates within the fuel element is needed. The method developed utilizes and interconnects the accuracy of the Monte Carlo Neutron/Photon (MCNP) method to calculate reaction rates for the important isotopes, and a time dependent depletion routine to calculate the temporal effects on isotope concentrations. This effort required the modification of MCNP itself to perform the additional task of accomplishing burnup calculations. The modified version called, MCNPBURN, evolved to be a general dual purpose code which can be used for standard calculations as well as for burn-up

  2. Thermal hydraulic analysis of 3 MW TRIGA research reactor of bangladesh considering different cycles of burnup

    International Nuclear Information System (INIS)

    Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core) was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt. (author)

  3. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

    Full Text Available Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt.

  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)

    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. 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.

  6. Technologies for manufacturing UO2 sintered pellets to fuel burnup extension

    International Nuclear Information System (INIS)

    The actual tendency all over the world is to manufacture fuel bundles capable to resist high burn-up. The factors affecting the burn-up increase are: the pellet-cladding mechanical interaction (PCMI), the oxidation and hydriding of the Zircaloy-4 sheath, the increase of internal pressure, stress corrosion cracking, Zircaloy-4 irradiation growth, fuel swelling. A way to increase fuel burn-up is to diminish the elements internal pressure by adequate UO2 fuel pellet structure (large grain or controlled closed porosity). In the large grain size UO2 pellets, fission gas release rate decreases and the elements internal pressure increase slowly. Similarly, in the UO2 sintered pellet with controlled closed porosity the fission gas accommodation is better and the elements internal pressure increases slowly. The paper presents a literature review related to the technologies and the methods for manufacturing UO2 sintered pellets to fuel burn-up extension. The flowsheets for large grains and controlled closed porosity UO2 sintered pellets obtained by Nb2O5 dopant respectively pores former addition in UO2 sinterable powder, pressing and sintering in H2 atmosphere are exposed. In the diagrams are presented the dependency of the main sintered pellet characteristics (pore radius distribution, pores volume, density, grains size) as function of the Nb2O5 dopant concentration, UO2 sinterable powder nature and sintering temperature. Other sintered pellets characteristics (electrical conductivity, Seebeck coefficient, high temperature molar heat capacity and thermomechanical properties) are presented. The technologies for sintered pellets manufacturing for RU, DUPIC, MOX fuel cycles are presented. A proposal related to fuel manufacturing from Uranium compound resulted in LWR spent fuel reprocessing is also given. (author)

  7. Analysis of the effect of UO2 high burnup microstructure on fission gas release

    International Nuclear Information System (INIS)

    This report deals with high-burnup phenomena with relevance to fission gas release from UO2 nuclear fuel. In particular, we study how the fission gas release is affected by local buildup of fissile plutonium isotopes and fission products at the fuel pellet periphery, with subsequent formation of a characteristic high-burnup rim zone micro-structure. An important aspect of these high-burnup effects is the degradation of fuel thermal conductivity, for which prevalent models are analysed and compared with respect to their theoretical bases and supporting experimental data. Moreover, the Halden IFA-429/519.9 high-burnup experiment is analysed by use of the FRAPCON3 computer code, into which modified and extended models for fission gas release are introduced. These models account for the change in Xe/Kr-ratio of produced and released fission gas with respect to time and space. In addition, several alternative correlations for fuel thermal conductivity are implemented, and their impact on calculated fission gas release is studied. The calculated fission gas release fraction in IFA-429/519.9 strongly depends on what correlation is used for the fuel thermal conductivity, since thermal release dominates over athermal release in this particular experiment. The conducted calculations show that athermal release processes account for less than 10% of the total gas release. However, athermal release from the fuel pellet rim zone is presumably underestimated by our models. This conclusion is corroborated by comparisons between measured and calculated Xe/Kr-ratios of the released fission gas

  8. Analysis of the effect of UO{sub 2} high burnup microstructure on fission gas release

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-10-01

    This report deals with high-burnup phenomena with relevance to fission gas release from UO{sub 2} nuclear fuel. In particular, we study how the fission gas release is affected by local buildup of fissile plutonium isotopes and fission products at the fuel pellet periphery, with subsequent formation of a characteristic high-burnup rim zone micro-structure. An important aspect of these high-burnup effects is the degradation of fuel thermal conductivity, for which prevalent models are analysed and compared with respect to their theoretical bases and supporting experimental data. Moreover, the Halden IFA-429/519.9 high-burnup experiment is analysed by use of the FRAPCON3 computer code, into which modified and extended models for fission gas release are introduced. These models account for the change in Xe/Kr-ratio of produced and released fission gas with respect to time and space. In addition, several alternative correlations for fuel thermal conductivity are implemented, and their impact on calculated fission gas release is studied. The calculated fission gas release fraction in IFA-429/519.9 strongly depends on what correlation is used for the fuel thermal conductivity, since thermal release dominates over athermal release in this particular experiment. The conducted calculations show that athermal release processes account for less than 10% of the total gas release. However, athermal release from the fuel pellet rim zone is presumably underestimated by our models. This conclusion is corroborated by comparisons between measured and calculated Xe/Kr-ratios of the released fission gas.

  9. Burnup credit calculations on long-term disposal

    International Nuclear Information System (INIS)

    One of the considered options for handling of irradiated nuclear fuel is the final disposal in some kind of repository. This necessitates the long-term investigation of subcriticality, heat production, public dose etc. NEA WPNCS Burnup Credit Expert Group defined a new benchmark to test the codes and data used for such problems. The effect of cooling time should be investigated. This implies that the decay data and not the cross sections influence the results. Composition of 4.5 % UO2 fuel with 50 MWd/kgU is given at the assembly removal from the core. Change of composition should be evaluated for 30 values of cooling time up to 1 million years. Keff should be evaluated with these compositions for a container housing 21 fuel assemblies. Initial concentration of 115 isotopes is given. For criticality calculations the usual 'burnup credit set' is used (14 actinides and 15 fission products). Results for additional isotopes is not presented now. The investigated fuel is 17 x 17 PWR UO2 type, with 25 guide tubes. The selected cooling times covers the time intervals of the usual handling procedures around the reactors (few years storing in storage pool, transport), interim storage (hundred years), and the long time scale of disposal up to 1 million years. Results: 1) For major actinides, ORIGEN and MULTICELL based keff results are practically identical up to 1000 years, far beyond the cooling times it was intended. 2) For actinides and fission products, the agreement is excellent up to 100 years, which covers the interim storage. 3) The difference of keff results about 0.02 at 1000 years. The reason is mainly the presence of Np-237, not considered in the previous case. It is produced from Am-241 by α-decay (432 years). Compositions calculated by ORIGEN and TIBSO results the same keff values for cooling times up to 1 million years. Changes in keff with cooling time have clear physical explanation. Compositions calculated by ORIGEN and MULTICELL results the same keff

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

    International Nuclear Information System (INIS)

    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.

  11. The comparison of recreative activities of 11-15 age group depending on different regions

    Directory of Open Access Journals (Sweden)

    Adem Pala

    2012-06-01

    Full Text Available The purpose of this study is to compare the recreative activities of 11-15 age group depending on different Regions. In this respect, 227 children from Southeastern part of Turkey and 262 children from Marmara region of Turkey participated in this study in a volunteered way. The questionnaire adapted from previous studies was implemented on these participants. Afterwards, the data were analyzed in terms of frequency, percentage and independent group of t-test statistics. Data were analyzed with the help of SPSS 15.0 and the significance rate was determined 0,05. Consequently, significantly different result were found (p<0,05 among participant from different regions, in their responses to the questions of “what they do in their leisure time, which recreative activities they regularly deal with, and which social and cultural activities they have joined in the past six months”. While the results found by the questions of “where they go to spend their spare time out of their school life and how many hour they allocate to recreative activities including sport” were completely insignificant (p>0,05. According to the findings of this research, 37,9 % of the participant from Southeastern part of Turkey chose “sports” option of the item of “what they do in their leisure time”, 53,3 % of participant within same region chose “football” option of the item of “which recreative activities they regularly deal with”. 51,5 % of the participant from Marmara region chose “sports” option of the former question: Additionally , 25,6 % of item chose “football”, 21,5 % of them chose “badminton” and 12,8 % of them chose “basketball”  options of the latter question.

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

    International Nuclear Information System (INIS)

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B4C 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, keff, 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 keff 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, keff 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

  13. Reactivity loss validation of high-burnup PWR fuels with pile-oscillation experiments in MINERVE

    International Nuclear Information System (INIS)

    The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a pressurized water reactor (PWR) between five and seven cycles, and also on the experimental validation of the spent fuel reactivity loss with burnup, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and the nuclear data responsible for the reactivity loss. This program also offers unique experimental data for fuels with a burnup reaching 85 GWd/tonne, as spent fuels in French PWRs have never exceeded 70 GWd/tonne up to now. The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first step, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists of the self-shielding of cross sections on the 281-energy-group SHEM mesh, followed by flux calculation by the method of characteristics in a two-dimensional 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 burnup of ∼75 GWd/tonne and within 4% for burnup of ∼85 GWd/tonne. (authors)

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

    International Nuclear Information System (INIS)

    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

  15. Divalent Metal- and High Mobility Group N Protein-Dependent Nucleosome Stability and Conformation

    Directory of Open Access Journals (Sweden)

    Michelle S. Ong

    2010-01-01

    Full Text Available High mobility group N proteins (HMGNs bind specifically to the nucleosome core and act as chromatin unfolding and activating factors. Using an all-Xenopus system, we found that HMGN1 and HMGN2 binding to nucleosomes results in distinct ion-dependent conformation and stability. HMGN2 association with nucleosome core particle or nucleosomal array in the presence of divalent metal triggers a reversible transition to a species with much reduced electrophoretic mobility, consistent with a less compact state of the nucleosome. Residues outside of the nucleosome binding domain are required for the activity, which is also displayed by an HMGN1 truncation product lacking part of the regulatory domain. In addition, thermal denaturation assays show that the presence of 1 mM Mg2+> or Ca2+ gives a reduction in nucleosome core terminus stability, which is further substantially diminished by the binding of HMGN2 or truncated HMGN1. Our findings emphasize the importance of divalent metals in nucleosome dynamics and suggest that the differential biological activities of HMGNs in chromatin activation may involve different conformational alterations and modulation of nucleosome core stability.

  16. Burnup instabilities in the full-core HTR model simulation

    International Nuclear Information System (INIS)

    Highlights: • We performed full-core burnup calculation coupled with Monte Carlo code. • Depletion instabilities have been detected for HTR system at high burnup. • We assess the stability of time step models in application to core calculation. • Discussion of the modeling factors related to burnup core simulation is presented. - Abstract: The phenomenon of numerical instabilities present in the Monte Carlo burnup calculations has been shown and explained by many authors using models of LWR, often simplified. Some theoretical considerations about origins of oscillations are very general, however it may be difficult to apply it easily to other models as a prediction of stability. Physics of HTR core differs significantly from the properties of light water system and the reliable extrapolation of the current numerical results is not possible. Moreover, most of the works concerning HTR burnup calculations put no emphasis on the spatial stability of the simulation and apply very long time steps. The awareness in this field of research seems to be not sufficient. In this paper, we focus on the demonstration of depletion instabilities in the simulations of HTR core dedicated for deep burnup of plutonium and minor actinides. We apply various methodology of time step implemented in advanced Continuous Energy Monte Carlo burnup code MCB version 5. Stability analysis is very rare for the full core calculations and the awareness of the oscillation’s problem is obligatory for the reliable modeling of a fuel cycle. In the summary of this work we systematize and discuss factors related to the stability of depletion and review available solutions

  17. Radionuclide Release from High Burnup Fuel

    International Nuclear Information System (INIS)

    In this paper we investigate the production, evolution and release of radioactive fission products in a light water reactor. The production of the nuclides is determined by the neutronics, their evolution in the fuel by local temperature and by the fuel microstructure and the rate of release is governed by the scenario and the properties of the microstructure where the nuclides reside. The problem combines fields of reactor physics, fuel behaviour analysis and accident analysis. Radionuclide evolution during fuel reactor life is also important for determination of instant release fraction of final repository analysis. The source term problem is investigated by literature study and simulations with reactor physics code Serpent as well as fuel performance code ENIGMA. The capabilities of severe accident management codes MELCOR and ASTEC for describing high burnup structure effects are reviewed. As the problem is multidisciplinary in nature the transfer of information between the codes is studied. While the combining of the different fields as they currently are is challenging, there are some possibilities to synergy. Using reactor physics tools capable of spatial discretization is necessary for determining the HBS inventory. Fuel performance studies can provide insight how the HBS should be modelled in severe accident codes, however the end effect is probably very small considering the energetic nature of the postulated accidents in these scenarios. Nuclide release in severe accidents is affected by fuel oxidation, which is not taken into account by ANSI/ANS-5.4 but could be important in some cases, and as such, following the example of severe accident models would benefit the development of fuel performance code models. (author)

  18. 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.…

  19. Using a Random Dependent Group Contingency to Increase On-Task Behaviors of High School Students with High Incidence Disabilities

    Science.gov (United States)

    Williamson, Brenda D.; Campbell-Whatley, Gloria D.; Lo, Ya-yu

    2009-01-01

    Group contingencies have the advantages of encouraging individual students to collectively feel responsible for appropriate and inappropriate classroom behaviors and have shown effectiveness in improving students' behavior. The purpose of this study was to investigate the effects of a random dependent group contingency on the on-task behaviors of…

  20. The impact of burn-up credit in criticality studies

    International Nuclear Information System (INIS)

    Nowadays optimization goes with everything. So French engineering firms try to demonstrate that fuel transport casks and storage pools are able to receive assemblies with higher 235U initial enrichments. Fuel Burnup distribution contributes to demonstrate it. This instruction has to elaborate a way to take credit of burnup effects on criticality safety designs. The calculation codes used are CESAR 4.21-APOLLO 1-MORET III. The assembly studied (UO2) is irradiated in a French Pressurized Water Reactor like EDF nuclear power reactor: PWR 1300 MWe, 17 x 17 array. Its initial enrichment in 235U equals 4.5%. The studies exposed in this report have evaluated the effects of: i) the 15 fission products considered in Burnup Credit (95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 133Cs, 143Nd, 145Nd, 147Sm, 149Sm, 150Sm, 151Sm, 152Sm, 153Eu, 155Gd), ii) the calculated abundances corrected or not by fixed factors, iii) the choice of one cross sections library used by CESAR 4.21, iu) the zone number elected in the axial burnup distribution zoning, u) the kind of cut applied on (regular/optimized). Two axial distribution profiles are studied: one with 44 GWd/t average burnup, the other with 20 GWd/t average burnup. The second one considers a shallow control rods insertion in the upper limit of the assembly. The results show a margin in reactivity about 0.045 with consideration of the 6 most absorbent fission products (103Rh, 133Cs, 143Nd, 149Sm, 152Sm, 155Gd), and about 0.06 for all Burnup Credit fission products whole. Those results have been calculated with an average burnup of 44 GWj/t. In a conservative approach, corrective factors must be apply on the abundance of some fission products. The cross sections library used by CESAR 4.21 (BBL 4) is sufficient and gives satisfactory results. The zoning of the assembly axial distribution burnup in 9 regular zones grants a satisfying calculation time/result precision compromise. (author)

  1. Application of burnup credit with partial boron credit to PWR spent fuel storage pools

    International Nuclear Information System (INIS)

    The outcome of performing a burnup credit criticality safety analysis of a PWR spent fuel storage pool is the determination of burnup credit loading curves BLC=BLC(e) for the spent fuel storage racks designed for burnup credit, cp. Reference. A burnup credit loading curve BLC=BLC(e) specifies the loading criterion by indicating the minimum burnup BLC(e) necessary for the fuel assembly with a specific initial enrichment e to be placed in storage racks designed for burnup credit. (orig.)

  2. Applications of ''candle'' burn-up strategy to several reactors

    International Nuclear Information System (INIS)

    The new burn-up strategy CANDLE is proposed, and the calculation procedure for its equilibrium state is presented. Using this strategy, the power shape does not change as time passes, and the excess reactivity and reactivity coefficient are constant during burn-up. No control mechanism for the burn-up reactivity is required, and power control is very easy. The reactor lifetime can be prolonged by elongating the core height. This burn-up strategy can be applied to several kinds of reactors whose maximum neutron multiplication factor changes from less than unity to more than unity, and then to less than unity. In the present paper it is applied to some fast reactors, thus requiring some fissile material such as plutonium for the nuclear ignition region of the core, but only natural uranium is required for the other region of the initial reactor and for succeeding reactors. The drift speed of the burning region for this reactor is about 4 cm/year, which is a preferable value for designing a long-life reactor. The average burn-up of the spent fuel is about 40%; that is, equivalent to 40% utilisation of the natural uranium without the reprocessing and enrichment. (author)

  3. The commercial and technological impact of high burnup

    International Nuclear Information System (INIS)

    Deregulation of electricity markets is driving prices downward. Consequently utilities continue to demand the minimization of electrical production costs. Fuel cycle cost savings are valued as a strong contributor, although directly representing only about one third of electricity generating costs. Burnups consistent with the current enrichment limit of 5 w/0 will be required. Significant progress has already been achieved by Siemens in meeting the demands of utilities for increased fuel burnup. The technological challenges imposed are mainly related to corrosion and hydrogen pickup of the clad, the properties of the fuel and the dimensional changes of the structure. Clad materials with increased corrosion resistance have been developed. The high burnup behaviour of the fuel has been extensively investigated and the decrease of thermal conductivity, the rim effect and the increase of fission gas release can be described, with good accuracy, in fuel rod computer codes. Advanced statistical design methods have been developed and introduced. In summary, most of the questions about the fuel operational behaviour and reliability in the high burnup range have been solved or the solutions are visible. The main licensing challenges for high burnup fuel are currently seen for accident condition analyses, especially for RIA and LOCA. (author)

  4. Burnup credit in nuclear waste transport: An economic analysis

    International Nuclear Information System (INIS)

    The US DOE is responsible for transporting nuclear spent fuel from commercial reactors to monitored retrievable storage (MRS) facilities and/or to repositories. Current plans call for approximately 110,000 metric tons uranium (MTU) to be transported over approximately 40 years beginning in 1998. Because of the large volume of spent fuel to be transported, new generations of spent fuel transportation casks are being planned. These casks will embody the latest technology and will be designated to accommodate the spent fuel in a way that maximizes the overall efficiency of the cask. In planning for the new generation of transport casks, the DOE is investigating the possibility of tailoring the cask design for the extent to which spent fuel has been used in the reactors, or, for spent fuel burnup. Granting design credit for burnup would allow one to fabricate casks with relatively larger capacities than would be possible otherwise. The remainder of the paper discusses the economic implications of using burnup credit in cask design, discusses the approach used in analyzing the economics of burnup credit, describes the results of the analysis, and offers some conclusions about the economic value of the burnup credit option

  5. Fuel cycle economical improvement by reaching high fuel burnup

    International Nuclear Information System (INIS)

    Improvements of fuel utilization in the light water reactors, burnup increase have led to a necessity to revise strategic approaches of the fuel cycle development. Different trends of the fuel cycle development are necessary to consider in accordance with the type of reactors used, the uranium market and other features that correspond to the nuclear and economic aspects of the fuel cycle. The fuel burnup step-by-step extension Program that successfully are being realized by the leading, firms - fuel manufacturers and the research centres allow to say that there are no serious technical obstacles for licensing in the near future of water cooling reactors fuel rod burnup (average) limit to 65-70 MWd/kgU and fuel assembly (average) limit to (60-65) MWd/kgU. The operating experience of Ukrainian NPPs with WWER-1000 is 130 reactor * years. At the beginning of 1999, a total quantity of the fuel FA discharged during all time of operation of 11 reactors was 5819 (110 fuel cycles). Economical improvement is reached by increase of fuel burn-up by using of some FA of 3 fuel cycles design in 4th fuel loading cycle. Fuel reliability is satisfactory. The further improvement of FA is necessary, that will allow to reduce the front-end fuel cycle cost (specific natural uranium expenditure), to reduce spent fuel amount and, respectively, the fuel cycle back end costs, and to increase burn-up of the fuel. (author)

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

    International Nuclear Information System (INIS)

    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 B4C 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

  7. Interdependent Group Contingency Management for Cocaine-Dependent Methadone Maintenance Patients

    Science.gov (United States)

    Kirby, Kimberly C.; Kerwin, MaryLouise E.; Carpenedo, Carolyn M.; Rosenwasser, Beth J.; Gardner, Robert S.

    2008-01-01

    Contingency management (CM) for drug abstinence has been applied to individuals independently even when delivered in groups. We developed a group CM intervention in which the behavior of a single, randomly selected, anonymous individual determined reinforcement delivery for the entire group. We also compared contingencies placed only on cocaine…

  8. Distance-Dependent Attractive and Repulsive Interactions of Bulky Alkyl Groups.

    Science.gov (United States)

    Hwang, Jungwun; Li, Ping; Smith, Mark D; Shimizu, Ken D

    2016-07-01

    The stabilizing and destabilizing effects of alkyl groups on an aromatic stacking interaction were experimentally measured in solution. The size (Me, Et, iPr, and tBu) and position (meta and para) of the alkyl groups were varied in a molecular balance model system designed to measure the strength of an intramolecular aromatic interaction. Opposite stability trends were observed for alkyl substituents at different positions on the aromatic rings. At the closer meta-position, smaller groups were stabilizing and larger groups were destabilizing. Conversely, at the farther para-position, the larger alkyl groups were systematically more stabilizing with the bulky tBu group forming the strongest stabilizing interaction. X-ray crystal structures showed that the stabilizing interactions of the small meta-alkyl and large para-alkyl groups were due to their similar distances and van der Waals contact areas with the edge of opposing aromatic ring. PMID:27159670

  9. Revisiting histidine-dependent acid phosphatases: a distinct group of tyrosine phosphatases

    OpenAIRE

    Veeramani, Suresh; Lee, Ming-Shyue; Lin, Ming-Fong

    2009-01-01

    Although classical protein tyrosine phosphatase (PTP) superfamily members are cysteine-dependent, emerging evidence shows that many acid phosphatases (AcPs) function as histidine-dependent PTPs in vivo. These AcPs dephosphorylate phospho-tyrosine substrates intracellularly and could have roles in development and disease. In contrast to cysteine-dependent PTPs, they utilize histidine, rather than cysteine, for substrate dephosphorylation. Structural analyses reveal that active site histidine, ...

  10. Mechanical Property Evaluation of High Burnup PHWR Fuel Clads

    International Nuclear Information System (INIS)

    Assurance of clad integrity is of vital importance for the safe and reliable extension of fuel burnup. In order to study the effect of extended burnup of 15,000 MW∙d/tU on the performance of Pressurised Heavy Water Reactor (PHWR) fuel bundles of 19-element design, a couple of bundles were irradiated in Indian PHWR. The tensile property of irradiated cladding from one such bundle was evaluated using the ring tension test method. Using a similar method, claddings of mixed oxide (MOX) fuel elements irradiated in the pressurized water loop (PWL) of CIRUS to a burnup of 10,000 MW∙d/THM were tested. The tests were carried out both at ambient temperature and at 300°C. The paper will describe the test procedure, results generated and discuss the findings. (author)

  11. A burnup credit calculation methodology for PWR spent fuel transportation

    International Nuclear Information System (INIS)

    A burnup credit calculation methodology for PWR spent fuel transportation has been developed and validated in CEA/Saclay. To perform the calculation, the spent fuel composition are first determined by the PEPIN-2 depletion analysis. Secondly the most important actinides and fission product poisons are automatically selected in PEPIN-2 according to the reactivity worth and the burnup for critically consideration. Then the 3D Monte Carlo critically code TRIMARAN-2 is used to examine the subcriticality. All the resonance self-shielded cross sections used in this calculation system are prepared with the APOLLO-2 lattice cell code. The burnup credit calculation methodology and related PWR spent fuel transportation benchmark results are reported and discussed. (authors)

  12. Study on the conservative factors for burnup credit criticality calculation

    International Nuclear Information System (INIS)

    When applies the burnup credit technology to perform criticality safety analysis for spent fuel storage or transportation problems, it is important for one to confirm that all the conditions adopted are adequate to cover the severest conditions that may encounter in the engineering applications. Taking the OECD/NEA burnup credit criticality benchmarks as sample problems, we study the effect of some important factors that may affect the conservatism of' the results for spent fuel system criticality safety analysis. Effects caused by different nuclides credit strategy, different cooling time and axial burnup profile are studied by use of the STARBUCS module of SCALE5. 1 software package, and related conclusions about the conservatism of these factors are drawn. (authors)

  13. High-burnup fuel and the impact on fuel management

    International Nuclear Information System (INIS)

    Competition in the electric utility industry has forced utilities to reduce cost. For a nuclear utility, this means a reduction of both the nuclear fuel cost and the operating and maintenance cost. To this extent, utilities are pursuing longer cycles. To reduce the nuclear fuel cost, utilities are trying to reduce batch size while increasing cycle length. Yankee Atomic Electric Company has performed a number of fuel cycle studies to optimize both batch size and cycle length; however, certain burnup-related constraints are encountered. As a result of these circumstances, longer fuel cycles make it increasingly difficult to simultaneously meet the burnup-related fuel design constraints and the technical specification limits. Longer cycles require fuel assemblies to operate for longer times at relatively high power. If utilities continue to pursue longer cycles to help reduce nuclear fuel cost, changes may need to be made to existing fuel burnup limits

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

    International Nuclear Information System (INIS)

    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)

  15. Burnup credit methodology validation against WWER experimental data

    International Nuclear Information System (INIS)

    A methodology for criticality safety analyses with burnup credit application has been developed for WWER spent fuel management facilities. This methodology is based on two worldwide used code systems: SCALE 4.4 for depletion and criticality calculations and NESSEL-NUKO - for depletion calculations. The methodology is in process of extensive validation for WWER applications. The depletion code systems NESSEL-NUKO and SCALE4.4 (control module SAS2H) have been validated on the basis of comparison with the calculated results obtained by other depletion codes for the CB2 Calculational Burnup Credit Benchmark. The validation of these code systems for WWER-440 and WWER-1000 spent fuel assembly depletion analysis based on comparisons with appropriate experimental data commenced last year. In this paper some results from burnup methodology validation against measured nuclide concentration given in the ISTC project 2670 for WWER-440 and from ORNL publication for WWER-1000 are presented. (authors)

  16. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

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

  17. Burnup credit implementation plan and preparation work at JAERI

    International Nuclear Information System (INIS)

    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)

  18. WWER fuel behaviour and characteristics at high burnup

    International Nuclear Information System (INIS)

    The increase of fuel burnup in fuel rods is a task that provides a considerable cost reduction of WWER fuel cycle in case of its solution. Investigations on fuel and cladding behaviour and change in fuel characteristics under irradiation are carried out in the Russian Federation for standard and as well as for experimental fuel rods to validate the reliable and safe operation of the fuel rods at high burnups. The paper presents the results of examinations on cracking, dimensional, structural and density changes of fuel pellets as well as the results of examination on corrosion and mechanical properties of WWER-440 and WWER-1000 fuel rod claddings. (author)

  19. Burnup measurements with the Los Alamos fork detector

    International Nuclear Information System (INIS)

    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

  20. Consequences of the increase of burnup on the fuel

    International Nuclear Information System (INIS)

    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

  1. Performance of fast reactor irradiated fueled emitters at goal burnup

    International Nuclear Information System (INIS)

    UO2-fueled W emitters were examined that had been irradiated to goal burnups of approximately 4 at.% at emitter surface temperatures to 1820 K in a fast reactor to establish their performance for use in thermionic reactors with power levels from tens of kilowatts to multimegawatts. The examinations provided first-time data on structural integrity, dimensional stability, component compatibility, and fuel and fission product behavior. The data are consistent with similar measurements at approximately 2 at.% burnup with the exception of one emitter which breached the W during irradiation

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

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D.J.; Neymoith, L.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

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

  3. 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 parametri

  4. Young Children Enforce Social Norms Selectively Depending on the Violator's Group Affiliation

    Science.gov (United States)

    Schmidt, Marco F. H.; Rakoczy, Hannes; Tomasello, Michael

    2012-01-01

    To become cooperative members of their cultural groups, developing children must follow their group's social norms. But young children are not just blind norm followers, they are also active norm enforcers, for example, protesting and correcting when someone plays a conventional game the "wrong" way. In two studies, we asked whether young children…

  5. 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. PMID:26839763

  6. Size Design of CdZnTe Detector Shield for Measuring Burnup of Spent Fuel

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>It is important to measure the burnup of spent fuel for nuclear safeguards, burnup credit and critical safety in spent-fuel reprocessing process. The purpose of this work is designing a portable device to

  7. Estimation of Parameters of the Rasch Model and Comparison of Groups in Presence of Locally Dependent Items.

    Science.gov (United States)

    Feddag, Mohand-Larbi; Blanchin, Myriam; Sébille, Véronique; Hardouin, Jean-Benoit

    2015-01-01

    Measurement specialists routinely assume examinee responses to test are independent of one another. However, previous research has shown that many tests contain item dependencies, and not accounting for these dependencies leads to misleading estimates of item and person parameters. In this paper, the marginal maximum likelihood estimation in Rasch model with the violation of the local independence is studied. The power of the Wald test on a group effect parameter on the latent traits in cross-sectional studies is examined under the local independence and the local item dependence assumptions. The different results are illustrated with simulation studies. PMID:26753222

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

    International Nuclear Information System (INIS)

    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)

  9. Validation of IRBURN calculation code system through burnup benchmark analysis

    International Nuclear Information System (INIS)

    Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.

  10. PWR fuel performance and burnup extension programme in Japan

    International Nuclear Information System (INIS)

    Since the first PWR nuclear power plant Mihama Unit 1 initiated commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts on improving the technology of PWRs. The results can already be seen by the significantly improved performance of the PWR plants now in operation. Mitsubishi Heavy Industries, Ltd supplied the nuclear fuel assemblies, which now amount to almost 5000. Although some trouble with fuel was experienced in the beginning, the progressive efforts made to improve the fuel design and manufacturing technology have resulted in the superior performance of Mitsubishi fuels. Since fuel of current design should comply with the limitation set in Japan for a maximum discharged fuel assembly average burnup of less than 39,000 MW·d/t, the maximum burnup is now around 37,000 MW·d/t. However, an increase in this burnup limitation has been strongly requested by Japanese utilities in order to make nuclear power more economic and thus more competitive with other power generation methods. A summary is given of the design improvements made on Mitsubishi fuel, as well as demonstration programmes of current design fuel to prove its superior reliability and to prepare the database for a future extension of burnup. (author)

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

    International Nuclear Information System (INIS)

    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

  12. Quantitative burnup determination: A comparison of different experimental methods

    International Nuclear Information System (INIS)

    The burn-up of nuclear fuel is defined as the energy produced per mass of fuel and, hence, is related to the inventory of fission products formed in the matrix of the fuel. It affects both neutron-physical and material properties. Therefore, it is essential to have methods available that allow a reliable determination of this important parameter. The burn-up is usually determined by measuring the content of an element that results from the fission process. The isotope 148Nd has proven to be an ideal monitor due to its chemical and neutron physical properties. On the other hand, 148Nd can only be determined by wet-chemistry methods, which means a rather costly and time consuming chemistry process. Another method using the sum of 145Nd and 146Nd is proposed. In case of very high burn-ups of U02 fuel and, especially, MOX fuel this method needs weighed yields for U and Pu to obtain a sufficient accuracy. Among the non-destructive spectrometric methods, the burn-up determination with 137Cs provides adequate results provided the gamma radiation detector is calibrated and self-attenuation effects of Cs together with measurement geometries are considered. (Author)

  13. Burn-up measurement of irradiated rock-like fuels

    International Nuclear Information System (INIS)

    In order to obtain burn-up data of plutonium rock-like (ROX) fuels irradiated at JRR-3M in JAERI, destructive chemical analysis of zirconia or thoria system ROX fuels was performed after development of a new dissolution method. The dissolution method and procedure have been established using simulated ROX fuel, which is applicable to the hot-cell handling. Specimens for destructive chemical analysis were obtained by applying the present method to irradiated ROX fuels in a hot-cell. Isotopic ratios of neodymium and plutonium were determined by mass-spectrometry using the isotope dilution procedure. Burn-up of the irradiated ROX fuels was calculated by the 148Nd procedure using measured data. The burn-ups of thoria and zirconia system fuels that irradiated same location in the capsule showed almost same values. For the ROX fuel containing thorium, 233U was also determined by the same techniques in order to evaluate the effect of burn-up of thorium. As the result, it was found that the fission of 233U was below 1% of total fission number and could be negligible. In addition, americium and curium were determined by alpha-spectrometry. These data, together with isotopic ratio of plutonium, are important data to analyze the irradiation behavior of plutonium. (author)

  14. Prediction of fission gas pressure from high burnup oxide fuel

    International Nuclear Information System (INIS)

    The ELESIM fuel performance code incorporates a fundamentally based treatment of the relevant physical processes affecting fission gas release. The fission gas release model treats fission gas diffusion, formation and subsequent interlinkage of intergranular bubbles, grain boundary storage of gas, grain growth and fuel swelling. The latter case considers the contributions of thermal expansion, densification, solid fission products, and gas bubbles. The effect of porosity on fuel thermal conductivity is taken into account. Previously we showed predictions of the gas release model agreed well with measured values for oxide fuel with burnups to about 300 MW.h/kg U. The applicability of the model to high burnup fuel is examined using examples from the literature. The fission gas release range considered is about 1-100% for burnups to 1000 MW.h/kg U in thermal reactor fuel and 2400 MW.h/kg U in fast reactor fuel. Predicted and measured releases are shown to be in good agreement, suggesting that the fundamental model is correct. In some models, empirical correction factors are required at high burnup to achieve agreement between predicted and measured release values; no such factor is required in ELESIM. (auth)

  15. Automated system for determining the burnup of spent nuclear fuel

    Directory of Open Access Journals (Sweden)

    Mokritskii V. A.

    2014-12-01

    Full Text Available The authors analyze their experience in application of semi-conductor detectors and development of a breadboard model of the monitoring system for spent nuclear fuel (SNF. Such system should use CdZnTe-detectors in which one-charging gathering conditions are realized. The proposed technique of real time SNF control during reloading technological operations is based on the obtained research results. Methods for determining the burnup of spent nuclear fuel based on measuring the characteristics of intrinsic radiation are covered in many papers, but those metods do not usually take into account that the nuclear fuel used during the operation has varying degrees of initial enrichment, or a new kind of fuel may be used. Besides, the known methods often do not fit well into the existing technology of fuel loading operations and are not suitable for operational control. Nuclear fuel monitoring (including burnup determination system in this research is based on the measurement of the spectrum of natural gamma-radiation of irradiated fuel assemblies (IFA, as from the point of view of minimizing the time spent, the measurement of IFA gamma spectra directly during fuel loading is optimal. It is the overload time that is regulated rather strictly, and burnup control operations should be coordinated with the schedule of the fuel loading. Therefore, the real time working capacity of the system should be chosen as the basic criterion when constructing the structure of such burnup control systems.

  16. Effects of cognitive and experiential group therapy on self-efficacy and perceptions of employability of chemically dependent women.

    Science.gov (United States)

    Washington, O

    1999-01-01

    This quasi-experimental study assessed effects of cognitive and experiential group therapy on self-efficacy and perceptions of employability for 52 chemically dependent adult women. The sample was 98% African American. Therapy consisted of six 90-min group sessions held twice weekly. The participants were pre- and posttested with the Self-Efficacy Scale (M. Sherer et al., 1982) and the Ghiselli Self-Description Inventory (E. E. Ghiselli, 1975). After the intervention, the cognitive group had significantly higher levels than the experiential group of social self-efficacy and need for self-actualization, an indicator of aspiration for employment. General self-efficacy and decisiveness, indicators of employability, significantly increased over time for both groups. Interventions to enhance people's belief in their ability to successfully perform tasks and control outcomes, promote personal growth, teach responsibility, and enhance self-awareness could be used to develop employability skills that reduce recidivism. PMID:10633639

  17. Handling requirements dependencies in agile projects: a focus group with agile software development practitioners

    NARCIS (Netherlands)

    Martakis, Aias; Daneva, Maya; Wieringa, R.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

  18. Need for higher fuel burnup at the Hatch Plant

    International Nuclear Information System (INIS)

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

  19. Need for higher fuel burnup at the Hatch Plant

    Energy Technology Data Exchange (ETDEWEB)

    Beckhman, J.T. [Georgia Power Co., Birmingham, AL (United States)

    1996-03-01

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

  20. Is in-group bias culture-dependent? A meta-analysis across 18 societies

    OpenAIRE

    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 intera...

  1. Monte Carlo burnup code acceleration with the correlated sampling method. Preliminary test on an UOX cell with TRIPOLI-4{sup R}

    Energy Technology Data Exchange (ETDEWEB)

    Dieudonne, C.; Dumonteil, E.; Malvagi, F.; Diop, C. M. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, Service d' Etude des Reacteurs et de Mathematiques Appliquees, DEN/DANS/DM2S/SERMA/LTSD, F91191 Gif-sur-Yvette cedex (France)

    2013-07-01

    For several years, Monte Carlo burnup/depletion codes have appeared, which couple a Monte Carlo code to simulate the neutron transport to a deterministic method that computes 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 time-expensive Monte Carlo solver called at each time step. Therefore, great improvements in term of calculation time could be expected if one could get rid of Monte Carlo transport sequences. For example, it may seem interesting to run an initial Monte Carlo simulation only once, for the first time/burnup step, and then to use the concentration perturbation capability of the Monte Carlo code to replace the other time/burnup steps (the different burnup steps are seen like perturbations of the concentrations of the initial burnup step). This paper presents some advantages and limitations of this technique and preliminary results in terms of speed up and figure of merit. Finally, we will detail different possible calculation scheme based on that method. (authors)

  2. The measurement of burn-up level in HTR-10

    International Nuclear Information System (INIS)

    Without shutting down the HTR-10, each fuel ball unloaded from the core must be measured. Judgment is made whether the desired burn-up level is reached. A fuel ball should be reloaded into the core when its burn-up level is less than 72,000 Mwd/tu. Since the measurement of burn-up level for a ball containing 0.9g 235U at most must be nondestructive, a γ spectroscopy method with high-resolution for the fission product 137Cs is typically chosen. However, because the HTR-10 is not provided with any kind of external calibrating source, it is impossible to achieve the goal that the accuracy of measurement is up to 2% by the above method. The method measuring burn-up levels without external source uses the ratio of 134Cs to 137Cs and gets bogged down in something unusual that both successful and failed examples have alternated in publications. The inner calibrating method is proposed in the paper. It is necessary to solve the following problems: a) The simple relationship between 134Cs/137Cs and burn-up level is held only in a specified range, but not for a spent ball. b) The migration must be ignored. c) How to deal with the neutron spectrum within HTR-10? The paper also introduces such useful method as neutron spectrum correction, picking out the graphite balls and extraction of 137Cs from γ spectra of reactor. The appropriate instrumentation that discriminates out the spent fuel balls and picks out the graphite balls is described. (author)

  3. 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.

  4. A Comparison of Independent, Interdependent, and Dependent Group Contingencies with Randomized Reinforcers to Increase Reading Fluency

    Science.gov (United States)

    Alric, Jolie M.; Bray, Melissa A.; Kehle, Thomas J.; Chafouleas, Sandra M.; Theodore, Lea A.

    2007-01-01

    Fluency, or the rate at which a student reads, is developed in the early stages of literacy and has been shown to correlate with comprehension. A myriad of interventions have been developed to increase fluency. Group contingencies are one of these that in particular have shown some positive effects on reading fluency. Advantages to using them are…

  5. 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…

  6. Behaviour of fuel rods of the second generation at high burnup WWER-440 fuel cycles. Aspects for attainment of burnup 70 MWd/kgU

    International Nuclear Information System (INIS)

    In this report an analysis of WWER-440 fuel of the second generation supplied by Russian JSC TVEL for high burnup fuel cycle is presented. The certificated code START-3 is applied to modeling of fuel rod operation parameters. Reliability of high-burnup fuel on the base of 5-6 year operation is demonstrated. Special attention is paid to aspects for attainment of burnup 70 MWd/kgU, including experimental and fuel modeling support and fuel operation experience

  7. Conditional Symmetry Groups of Nonlinear Diffusion Equations with x-Dependent Convection and Absorption

    Institute of Scientific and Technical Information of China (English)

    QU Chang-Zheng; ZHANG Shun-Li

    2004-01-01

    The generalized conditionalsymmetry and sign-invariant approaches are developed to study the nonlinear diffusion equations with x-dependent convection and source terms. We obtain conditions under which the equations admit the second-order generalized conditional symmetries and the first-order sign-invariants on the solutions. Several types of different generalized conditional symmetries and first-order sign-invariants for the equations with diffusion of power law are obtained. Exact solutions to the resulting equations are constructed.

  8. 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)

  9. Addition of the sulfhydryl group (-SH) to the PPR78 model (predictive 1978, Peng-Robinson EOS with temperature dependent kij calculated through a group contribution method)

    International Nuclear Information System (INIS)

    In 2004, we started to develop a group contribution method aimed at estimating the temperature dependent binary interaction parameters (kij(T)) for the widely used Peng-Robinson equation of state (EOS). This model was called PPR78 (predictive 1978, Peng-Robinson EOS) because it relies on the Peng-Robinson EOS as published by Peng and Robinson in 1978 and because the addition of a group contribution method to estimate the kij makes it predictive. In our previous papers, 14 groups were defined: CH3, CH2, CH, C, CH4 (methane), C2H6 (ethane), CHaro, Caro, Cfusedaromaticrings, CH2,cyclic, CHcyclic=Ccyclic, CO2, N2, and H2S. It was thus possible to estimate the kij for any mixture containing alkanes, aromatics, naphthenes, CO2, N2, and H2S whatever the temperature. In this study, the PPR78 model is extended to systems containing thiols (also called mercaptans). To do so, the sulfhydryl group: -SH was added

  10. Effect of Self-Shielding on Burn-Up Calculation of ETRR-2 Reactor

    International Nuclear Information System (INIS)

    There exist two approaches for burn-up calculation. The first on is to use cell parameters generated using cell calculation code at different degrees of burn-up. The other is to use microscopic cross sections with self-shielding in order to compensate for the variation of spectrum at different degree of burn-up. The effect of using different forms of self-shielding factors on burn-up calculation for ETRR-2 reactor has been determined. The results of the two approaches are inter-compared up to 50% burn-up

  11. Revaluation on measured burnup values of fuel assemblies by post-irradiation experiments at BWR plants

    International Nuclear Information System (INIS)

    Fuel composition data for 8x8 UO2, Tsuruga MOX and 9x9-A type UO2 fuel assemblies irradiated in BWR plants were measured. Burnup values for measured fuels based on Nd-148 method were revaluated. In this report, Nd-148 fission yield and energy per fission obtained by burnup analyses for measured fuels were applied and fuel composition data for the measured fuel assemblies were revised. Furthermore, the adequacies of revaluated burnup values were verified through the comparison with burnup values calculated by the burnup analyses for the measured fuel assemblies. (author)

  12. Investigation of several methods to set burnup for criticality safety assessment of spent fuel transport casks

    International Nuclear Information System (INIS)

    Several currently available methods to set burnup for depletion calculation are reviewed and discussed about its adequacy for criticality safety assessment of spent fuel (SF) transport casks by taking burnup credit (BC) into accounts. Various errors associated with BC criticality analyses are evaluated and converted to equivalent burnup to compare each other. Methods are proposed to use some reduced burnups equivalent to compensation of these associated errors. Effects of assumption of axial burnup distribution on criticality calculation and irradiation history parameter variation on depletion calculation are evaluated with OECD/NEA BC international benchmark data. (author)

  13. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    International Nuclear Information System (INIS)

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO2 fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

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

    International Nuclear Information System (INIS)

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

  15. 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...

  16. Burn-up credit criticality benchmark. Phase 4-A: reactivity prediction calculations for infinite arrays of PWR MOX fuel pin cells

    International Nuclear Information System (INIS)

    The OECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs. The report summarises and analyses the solutions to the specified exercises provided by 37 contributors from 10 countries. The exercises were based upon the calculation of infinite PWR fuel pin cell reactivity for fresh and irradiated MOX fuels with various MOX compositions, burn-ups and cooling times. In addition, several representations of the MOX fuel assembly were tested in order to check various levels of approximations commonly used in reactor physics calculations. (authors)

  17. Continuation of the WWER burnup credit benchmark: evaluation of CB1 results, overview of CB2 results to date, and specification of CB3

    International Nuclear Information System (INIS)

    A calculational benchmark focused on WWER-440 burnup credit, simular to that of the OECD/NEA/NSC Burnup Credit Criticality Benchmark Working Group, was proposed on the 96'AER Symposium. Its first part, CB1, was specified there whereas the second part, CB2, was specified a year later, on 97'AER Symposium in Zittau. This paper brings a final statistical evaluation of CB1 results and summarizes the CB2 results obtained to date. Further, the effect of an axial burnup profile of WWER-440 spent fuel on criticality ('end effect') is proposed to be studied in the CB3 benchmark problem of an infinite array of WWER-440 spent fuel rods as specified in the paper. (Authors)

  18. Temperature-Dependent Structure-Energy Changes in Crystals of Compounds with Poly(hydroxymethyl) Grouping

    Science.gov (United States)

    Golovina, N. I.; Raevskii, A. V.; Fedorov, B. S.; Chukanov, N. V.; Shilov, G. V.; Leonova, L. S.; Tarasov, V. P.; Erofeev, L. N.

    2002-03-01

    Crystals of tris(hydroxymethyl)nitromethane (1) and tris(hydroxymethyl)aminomethane (2) were prepared and grown at room temperature. X-ray analysis was used to study the structure of crystals 1 and 2 at room temperature; the X-ray diffraction method was applied to investigate polycrystalline samples during a temperature rise up to the phase transition into the plastic phase. Phase transitions in separate crystals 1 and 2 were observed in a hot stage under an optical microscope. Calorimetric study of the crystal temperature behavior and the phase transition features including melting were carried out. By IR spectroscopy the temperature relations of the bonds of symmetric N-O stretching vibrations of nitro groups and stretching vibrations of OH groups redistribution in crystals of 1 were investigated. In crystals of 2 the behavior of stretching vibration bands of O-H groups was studied at room temperature. In the temperature interval including phase transition, data on structure-dynamic rearrangements in the crystal lattice of compounds 1 and 2 were obtained by the NMR pulse method in the solid phase using relaxational free induction decay of protons. The proton conductivity was found and its temperature parameters were determined for both compounds in the plastic state.

  19. 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. PMID:26674073

  20. Analysis of neodymium 148 in order to determin of nuclear fuel burnup

    International Nuclear Information System (INIS)

    To determine the degree of the nuclear fuel burnup experiments were conducted to introduce improvements in the mass-spectrometric study of neodymium-148 by the method of isotopic dilution with Nd-150 taken as a diluent. The separation of neodymium out of the mixture of the fission products and uranium was carried out in two stages. In the first stage a group of rare earth elements was isolated on the Vofatit SBV anionite in the mixture of nitric acid and methanol. The second stage involved the separation of the rare earth group on the Vofatite KPS cationite with the aid of the complexing agent of α-hydroxy-isobutyric acid. To identify the neodymium fraction, the traces of americium-241 were added at elution. The possibilities of the above analytical method are examplified by the isolation of neodymium out of the burned-up fuel of type EK-10. The isotopic ratios were determined by the spectroscopic method to the accuracy of +-1.2%. A highly enriched compound of neodymium-150 was used as a diluent. The factors are discussed affecting the degree of the burnup obtained by this method

  1. Renormalization group evolution of the Standard Model dimension six operators II: Yukawa dependence

    International Nuclear Information System (INIS)

    We calculate the complete order y2 and y4 terms of the 59×59 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 λ, λ2 and λy2 depending on the Standard Model Higgs self-coupling λ 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

  2. Finite difference solution of the time dependent neutron group diffusion equations

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

  4. Nonequilibrium frequency-dependent noise through a quantum dot: A real-time functional renormalization group approach

    OpenAIRE

    Moca C.P.; Simon P; Chung C.H.; Zarand G.

    2011-01-01

    We construct a real time current-conserving functional renormalization group (RG) scheme on the Keldysh contour to study frequency-dependent transport and noise through a quantum dot in the local moment regime. We find that the current vertex develops a non-trivial non-local structure in time, governed by a new set of RG equations. Solving these RG equations, we compute the complete frequency and temperature-dependence of the noise spectrum. For voltages large compared to the Kondo temperatur...

  5. Effects of burn-up, recovery efficiency and waste form on the environmental impact of fusion-fission transmutation systems

    International Nuclear Information System (INIS)

    The effects of fuel cycle parameters on nuclear waste environmental impact are analyzed for an advanced system that includes a Fusion-Fission Hybrid reactor. The system aims at reduction of the long-term radiotoxicity of waste by transmuting highly radiotoxic transuranics. However, the radiological risk of the system is measured by annual doses to the public, which are controlled by reactor operations, fuel cycle processes, waste treatment processes, and design of geological repositories. In this study, the waste environmental impact for a fuel cycle with a Fusion-Fission transmutation is analyzed as a function of three different parameters: burn-up, recovery efficiency and waste form durability for two different geological repositories, one with low actinide solubility and the other with high solubility. It is found that burn-up and recovery efficiency effects on environmental impact strongly depend on repository conditions, while the most influential parameter is found to be the durability of the waste form. (author)

  6. Chemical modification and pH dependence of kinetic parameters to identify functional groups in a glucosyltransferase from Strep. Mutans

    International Nuclear Information System (INIS)

    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 14C 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. Efficacy of treatment in an opioid -dependent population group using the Maudsley Addiction Profile (MAP) tool.

    Science.gov (United States)

    Collins, Ruth; Boggs, Bob; Taggart, Noel; Kelly, Martin; Drillington, Aileen; Swanton, Ivy; Patterson, Diane

    2009-01-01

    A pilot study was performed to assess the effectiveness of treatment in an opioid dependent population using the Maudsley Addiction Profile (MAP) tool1.The primary outcome of the study was to assess if treatment had an effect on 1. Substance use (quantity and frequency of use), 2. Health risk behaviour (injecting and sharing injecting equipment), 3. Health symptoms (physical and psychological) and 4. Personal /Social functioning (relationships, employment and crime). A secondary outcome was also sought.The study took place in 2007 in an inner city Belfast hospital specialising in the treatment of addiction, over a two month period. Fifteen patients, all opioid dependent and receiving outpatient community treatment, were interviewed at baseline (prior to the commencement of treatment) and at eight weeks follow up.Three patients were lost to follow up. Two patients stopped using altogether. Of the remaining patients, improvements were seen in most areas. There was a decrease in the use of heroin (71.28%), cocaine (99.72%), crack cocaine (100%), cannabis (99.94%) and alcohol (33.17%). There was a reduction in injecting behaviour (60.93%). Improvements were observed in health with a reduction in physical (41.35%) and psychological (35%) symptoms. Overall personal and social functioning improved regarding interactions with family and friends. A reduction in crime was also observed (75%).Opinions and views of staff involved in the study were generally positive.This patient population presents with multiple and complex needs. Effective treatment needs to address these needs and not just drug addiction alone. The Maudsley Addiction Profile tool highlights this. PMID:19252726

  8. Burn-up credit criticality benchmark. Phase 4-B: results and analysis of MOX fuel depletion calculations

    International Nuclear Information System (INIS)

    The DECD/NEA Expert Group on Burn-up Credit was established in 1991 to address scientific and technical issues connected with the use of burn-up credit in nuclear fuel cycle operations. Following the completion of six benchmark exercises with uranium oxide (UOX) fuels irradiated in pressurised water reactors (PWRs) and boiling water reactors (BWRs), the present report concerns mixed uranium and plutonium oxide (MOX) fuels irradiated in PWRs. The exercises consisted of inventory calculations of MOX fuels for two initial plutonium compositions. The depletion calculations were carried out using three representations of the MOX assemblies and their interface with UOX assemblies. This enabled the investigation of the spatial and spectral effects during the irradiation of the MOX fuels. (author)

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

    International Nuclear Information System (INIS)

    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

  10. Taking burnup credit into account in criticality studies: the situation as it is now and the prospect for the future

    International Nuclear Information System (INIS)

    As enrichment of the fuel has become higher than the limits used at the designing stages, it seemed necessary to consider fuel depletion during irradiation to guarantee the criticality safety for relatively high enriched fuels transportation, storage or reprocessing. This burnup credit will make it possible to use the devices for spent fuels which are initially relatively high enriched. For that purpose, a method was developed considering: (i) partial Uranium-and-Plutonium burnup credit in the criticality studies, and (ii) a conservative assumption concerning the axial profile; this actinides-only method was supported by an experimental program called HTC. The method was accepted by the French Safety Authority. Moreover, in order to reduce again the calculated values of the reactivity for irradiated fuels, a French working group was set up in 1997 to define a conservative method which enables industrial companies to take burnup credit into account with some of the fission products and using a more precise profile. The work of this group has been divided into four tasks related to: the determination of (i) the composition of the fuel, (ii) a conservative profile, (iii) a conservative irradiation history, and (iv) the calculation scheme. This work is also supported by experimental programs related to the validation of the fission products effects, in terms of reactivity

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

    International Nuclear Information System (INIS)

    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 UO2 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

  12. A deep-seated mechanism for cycle-dependent sunspot group tilt angles

    Science.gov (United States)

    Isik, Emre

    2016-07-01

    The cycle-averaged tilt angle of sunspot groups is an important quantity in determining the magnetic flux diffusing across the equator, which is highly correlated with the strength of the next cycle. This quantity has recently been reported to be anti-correlated with the strength of the solar cycle. I suggest that a deep-seated thermodynamic cycle can be responsible for the observed correlation. Motivated by helioseismic indications, I calculate the effect of cooling of the convective overshoot region on the stability and dynamics of thin, unstable flux tubes. I find that only 5-20 K of cooling in the layer can explain the observed range of tilt angle fluctuations among different cycles. This mechanism can play a role in the nonlinear saturation and amplitude fluctuations of the solar dynamo.

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

    International Nuclear Information System (INIS)

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

  14. A model describing the pressure dependence of the band gap energy for the group III-V semiconductors

    Science.gov (United States)

    Zhao, Chuan-Zhen; Wei, Tong; Sun, Xiao-Dong; Wang, Sha-Sha; Lu, Ke-Qing

    2016-08-01

    A model describing the pressure dependence of the band gap energy for the group III-V semiconductors has been developed. It is found that the model describes the pressure dependence of the band gap energy very well. It is also found that, although the pressure dependence of the band gap energy for both the conventional III-V semiconductors and the dilute nitride alloys can be described well by the model in this work, the physical mechanisms for them are different. In addition, the influence of the nonlinear compression of the lattice on the band gap energy is smaller than that of the coupling interaction between the N level and the conduction band minimum of the host material.

  15. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Su, Bingjing; Hawari, Ayman, I.

    2004-03-30

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from {approx} {+-}40% at beginning of life to {approx} {+-}10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this

  16. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    International Nuclear Information System (INIS)

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from ∼ ±40% at beginning of life to ∼ ±10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this case, a self

  17. Tritium Burn-up Depth and Tritium Break-Even Time

    Institute of Scientific and Technical Information of China (English)

    LI Cheng-Yue; DENG Bai-Quan; HUANG Jin-Hua; YAN Jian-Cheng

    2006-01-01

    @@ Similarly to but quite different from the xenon poisoning effects resulting from fission-produced iodine during the restart-up process of a fission reactor, we introduce a completely new concept of the tritium burn-up depth and tritium break-even time in the fusion energy research area. To show what the least required amount of tritium storage is used to start up a fusion reactor and how long a time the fusion reactor needs to be operated for achieving the tritium break-even during the initial start-up phase due to the finite tritium breeding time that is dependent on the tritium breeder, specific structure of breeding zone, layout of coolant flow pipe, tritium recovery scheme, extraction process, the tritium retention of reactor components, unrecoverable tritium fraction in breeder, leakage to the inertial gas container, and the natural decay etc., we describe this new phenomenon and answer this problem by setting up and by solving a set of equations, which express a dynamic subsystem model of the tritium inventory evolution in a fusion experimental breeder (FEB). It is found that the tritium burn-up depth is 317g and the tritium break-even time is approximately 240 full power days for FEB designed detail configuration and it is also found that after one-year operation, the tritium storage reaches 1.18kg that is more than theleast required amount of tritium storage to start up three of FEB-like fusion reactors.

  18. Interaction of PAMAM dendrimers with bovine insulin depends on nanoparticle end-groups

    Energy Technology Data Exchange (ETDEWEB)

    Nowacka, Olga; Milowska, Katarzyna, E-mail: milowska@biol.uni.lodz.pl; Bryszewska, Maria

    2015-06-15

    We have looked at the interactions between polyamidoamine (PAMAM) dendrimers with different terminal groups (−COOH, −NH{sub 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.

  19. Interaction of PAMAM dendrimers with bovine insulin depends on nanoparticle end-groups

    International Nuclear Information System (INIS)

    We have looked at the interactions between polyamidoamine (PAMAM) dendrimers with different terminal groups (−COOH, −NH2, −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

  20. 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

    2014-05-01

    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. Their group models, which were represented by discourse and blood circulation diagrams, were analyzed for the development of the framework that informed the model evaluation levels and epistemological features. The model evaluation levels were categorized into levels one to four based on the following evaluation criteria: no evaluation, authoritative sources, superficial criteria, and more comprehensive criteria. The qualities of group models varied with the criteria of model evaluation. While students who used authoritative sources for evaluating the group model appeared to have an absolutist epistemology, students who evaluated according to the superficial criteria and more comprehensive criteria appeared to have an evaluative epistemology. Furthermore, groups with Level four showed a chain reaction of cognitive reasoning during the modeling practice concerning practical epistemology. The findings have implications for science teachers and education researchers who want to understand the context for developing students' practical epistemologies.

  1. Multivariate and repeated measures (MRM): A new toolbox for dependent and multimodal group-level neuroimaging data

    Science.gov (United States)

    McFarquhar, Martyn; McKie, Shane; Emsley, Richard; Suckling, John; Elliott, Rebecca; Williams, Stephen

    2016-01-01

    Repeated measurements and multimodal data are common in neuroimaging research. Despite this, conventional approaches to group level analysis ignore these repeated measurements in favour of multiple between-subject models using contrasts of interest. This approach has a number of drawbacks as certain designs and comparisons of interest are either not possible or complex to implement. Unfortunately, even when attempting to analyse group level data within a repeated-measures framework, the methods implemented in popular software packages make potentially unrealistic assumptions about the covariance structure across the brain. In this paper, we describe how this issue can be addressed in a simple and efficient manner using the multivariate form of the familiar general linear model (GLM), as implemented in a new MATLAB toolbox. This multivariate framework is discussed, paying particular attention to methods of inference by permutation. Comparisons with existing approaches and software packages for dependent group-level neuroimaging data are made. We also demonstrate how this method is easily adapted for dependency at the group level when multiple modalities of imaging are collected from the same individuals. Follow-up of these multimodal models using linear discriminant functions (LDA) is also discussed, with applications to future studies wishing to integrate multiple scanning techniques into investigating populations of interest. PMID:26921716

  2. Multivariate and repeated measures (MRM): A new toolbox for dependent and multimodal group-level neuroimaging data.

    Science.gov (United States)

    McFarquhar, Martyn; McKie, Shane; Emsley, Richard; Suckling, John; Elliott, Rebecca; Williams, Stephen

    2016-05-15

    Repeated measurements and multimodal data are common in neuroimaging research. Despite this, conventional approaches to group level analysis ignore these repeated measurements in favour of multiple between-subject models using contrasts of interest. This approach has a number of drawbacks as certain designs and comparisons of interest are either not possible or complex to implement. Unfortunately, even when attempting to analyse group level data within a repeated-measures framework, the methods implemented in popular software packages make potentially unrealistic assumptions about the covariance structure across the brain. In this paper, we describe how this issue can be addressed in a simple and efficient manner using the multivariate form of the familiar general linear model (GLM), as implemented in a new MATLAB toolbox. This multivariate framework is discussed, paying particular attention to methods of inference by permutation. Comparisons with existing approaches and software packages for dependent group-level neuroimaging data are made. We also demonstrate how this method is easily adapted for dependency at the group level when multiple modalities of imaging are collected from the same individuals. Follow-up of these multimodal models using linear discriminant functions (LDA) is also discussed, with applications to future studies wishing to integrate multiple scanning techniques into investigating populations of interest. PMID:26921716

  3. Actinide-only burnup credit for spent fuel transport

    International Nuclear Information System (INIS)

    A conservative methodology is described that would allow taking credit for burn up in the criticality safety analysis of spent nuclear fuel packages. Requirements for its implementation include isotopic and criticality validation, generation of package loading criteria using limiting parameters, and assembly burn up verification by measurement. The method allows credit for the changes in the 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am concentrations with burnup. No credit for fission product neutron absorbers is taken. Analyses are included regarding the methodology's financial benefits and conservative margin. It is estimated that the proposed actinide-only burnup credit methodology would save 20% of the transport costs. Nevertheless, the methodology includes a substantial margin. Conservatism due to the isotopic correction factors, limiting modelling parameters, limiting axial profiles and exclusion of the fission products ranges from 10 to 25% k. (author)

  4. Advances In Burnup Credit Criticality Safety Analysis Methods And Applications

    International Nuclear Information System (INIS)

    An International Workshop on “Advances in Applications of Burnup Credit for Spent Fuel Storage, Transport, Reprocessing, and Disposition” organized by the Nuclear Safety Council of Spain (CSN) in cooperation with the International Atomic Energy Agency (IAEA) was held at Córdoba, Spain, on October 27– 30, 2009. The objectives of this workshop were to identify the benefits that accrue from recent improvements of the burnup credit (BUC) analysis methodologies, to analyze the implications of applying improved BUC methodologies, focusing on both the safety-related and operational aspects, and to foster the exchange of international experience in licensing and implementation of BUC applications. In the paper on hand the attention is focused on the improvements of BUC analysis methodologies. (author)

  5. Burnup calculations using serpent code in accelerator driven thorium reactors

    International Nuclear Information System (INIS)

    In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232Th and mixed 233U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)

  6. Burnup calculations using serpent code in accelerator driven thorium reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, M.E.; Agar, O. [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Physics Dept.; Yigit, M. [Aksaray Univ. (Turkey). Physics Dept.

    2013-07-15

    In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed {sup 232}Th and mixed {sup 233}U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)

  7. Testing of a burnup measuring prototype. Final report

    International Nuclear Information System (INIS)

    The analyses of gamma spectroscopy measurements of spherical fuel elements are reported, which have been performed in order to prove the feasibility of burnup measurement by way of Cs-137 spectroscopy. The detailed analysis and evaluation of the FRJ2-KA2) irradiation experiment carried out in the DIDO reactor supplied clear evidence that it is possible to measure an HTR-Modul fuel pebble with a target burnup ob 80,000 MWd/t after a decay time of 55 h, within a period of 10 seconds and with an accuracy of <5%, with 1 σ confidence, and that there is no need for developing measuring instruments with a higher counting rate. There are commercial peak unfolding programs available. (orig./HP)

  8. Transient behaviour of high burnup fuel. Status report

    International Nuclear Information System (INIS)

    This Status Report is a follow-on to the CSNI Specialist Meeting on Transient Behaviour of High Burnup Fuel which was held in Cadarache, France, from September 12. to 14., 1995. The Status Report identifies the needs and rationale for any further work to better understand the transient behaviour of high burnup fuel. The different options to perform that work, from analytical to experimental activities, and discussion on the potential benefits of performing new integral tests are also addressed. A brief description of the major on-going and short-term planned activities in this field is included as additional information. The main conclusions from this effort are highlighted. (K.A.)

  9. OREST - The hammer-origen burnup program system

    International Nuclear Information System (INIS)

    Reliable prediction of the characteristics of irradiated light water reactor fuels (e.g., afterheat power, neutron and gamma radiation sources, final uranium and plutonium contents) is needed for many aspects of the nuclear fuel cycle. Two main problems must be solved: the simulation of all isotopic nuclear reactions and the simulation of neutron fluxes setting the reactions in motion. In state-of-the-art computer techniques, a combination of specialized codes for lattice cell and burnup calculations is preferred to solve these cross-linked problems in time or burnup step approximation. In the program system OREST, developed for official and commercial tasks in the Federal Republic of Germany nuclear fuel cycle, the well-known codes HAMMER and ORIGEN and directly coupled with a fuel rod temperature module

  10. Analysis of the burnup credit benchmark with an updated WIMS-D Library

    International Nuclear Information System (INIS)

    The OECD/NEA Burnup Credit Benchmark was analyzed with the WIMSD5B code using a fully updated library based on ENDF/B-VI Revision 5 data. Parts-1A and 1B were considered. The criticality prediction tested in Part-1A was in very good agreement with the reference result. A slight trend to overestimate the absorption rate by the fission products was noted, which can be explained by spectral effects resulting from the coarseness of the WIMS-D 69-group energy grid. The isotopic composition prediction tested in Part-1B was within the uncertainty interval of the reference results, except for 109 Ag at lower burnup and 155 Gd in all the cases. For 109 Ag the cause of the discrepancy was the use of old fission yield data in generating the reference solution. Similarly for 155 Gd the difference was due to old 155 Eu capture cross sections. Compared to the measurements, a serious underprediction of Sm isotopes is observed. This could be due to problems in the measured values or in the nuclear data of Sm precursors. We conclude that our processing methods do not introduce significant errors to the basic nuclear data. Care should be taken in the interpretation of the reference average benchmark solution due to a possible bias towards the ENDF/B-V evaluated nuclear data files

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

    International Nuclear Information System (INIS)

    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)

  12. 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.

  13. Value of 236U to actinide-only burnup credit

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) submitted a topical report to the US Nuclear Regulatory Commission (NRC) in May 1995 in order to gain approval of a method for criticality analysis of transport packages that takes account for the change in actinide isotopes with burnup [pressurized water reactors (PWRs) only]. Historically, the NRC has conservatively assumed that the fuel was in its initial conditions (without any burnable absorbers). In order to permit credit for the changes in actinide content, the NRC has required validation of the depletion and criticality codes for spent nuclear fuel, justification of conservative depletion modeling, and finally confirmation measurements before loading. The NRC requested additional information on March 22, 1996. The DOE responded by a revision of the topical report in May 1997. The NRC again responded with another set of requests of additional information in April 1998. In that set of questions, the NRC challenged the use of 236U in burnup credit. Uranium-236 is not found in any significant amount in any available critical experiments. The authors explore the value of 236U to actinide-only burnup credit

  14. Investigation of Burnup Credit Issues in BWR Fuel

    International Nuclear Information System (INIS)

    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

  15. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    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)

  16. Evolution of the ELESTRES code for application to extended burnups

    International Nuclear Information System (INIS)

    The computer code ELESTRES is frequently used at Atomic Energy of Canada Limited to assess the integrity of CANDU fuel under normal operating conditions. The code also provides initial conditions for evaluating fuel behaviour during high-temperature transients. This paper describes recent improvements in the code in the areas of pellet expansion and of fission gas release. Both of these are very important considerations in ensuring fuel integrity at extended burnups. Firstly, in calculations of pellet expansion, the code now accounts for the effect of thermal stresses on the volume of gas bubbles at the boundaries of UO2 grains. This has a major influence on the expansion of the pellet during power-ramps. Secondly, comparisons with data showed that the previous fission gas package significantly underpredicted the fission gas release at high burnups. This package has now been improved via modifications to the following modules: distance between neighbouring bubbles on grain boundaries; diffusivity; and thermal conductivity. The predictions of the revised version of the code show reasonable agreement with measurements of ridge strains and of fission gas release. An illustrative example demonstrates that the code can be used to identify a fuel design that would: reduce the sheath stresses at circumferential ridges by a factor of 2-10; and keep the gas pressure at very high burnups to below the coolant pressure

  17. The REBUS experimental programme for burn-up credit

    International Nuclear Information System (INIS)

    An international programme called REBUS for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Centre SCK·CEN and Belgonucleaire with the support of EdF and IRSN from France and VGB, representing German nuclear utilities and NUPEC, representing the Japanese industry. Recently also ORNL from the U.S. jointed the programme. The programme aims to establish a neutronic benchmark for reactor physics codes in order to qualify the codes for calculations of the burn-up credit. The benchmark exercise investigate the following fuel types with associated burn-up: reference fresh 3.3% enriched UO2 fuel, fresh commercial PWR UO2 fuel and irradiated commercial PWR UO2 fuel (54 GWd/tM), fresh PWR MOX fuel and irradiated PWR MOX fuel (20 GWd/tM). The experiments on the three configurations with fresh fuel have been completed. The experiments show a good agreement between calculation and experiments for the different measured parameters: critical water level, reactivity effect of the water level and fission-rate and flux distributions. In 2003 the irradiated BR3 MOX fuel bundle was loaded into the VENUS reactor and the associated experimental programme was carried out. The reactivity measurements in this configuration with irradiated fuel show a good agreement between experimental and preliminary calculated values. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 Gd2O3 mixed with fuel or with inert element like Al2O3. 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 Gd2O3 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

  20. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    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. Pore pressure calculation of the UO2 high burnup structure

    International Nuclear Information System (INIS)

    Highlights: • Pore pressure is calculated based on local burnup, density and porosity. • Ronchi's equations of state are used instead of van der Waals’ equation. • Pore pressure increases as HBS transformation begins and then stays constant. • A best approximated parameter used for pore pressure calculation is recommended. -- Abstract: UO2 high burnup structure has an important impact on fuel behavior, especially in case of reactivity initiated accident (RIA). Pore relaxation enhances local fuel swelling and puts additional load to the fuel cladding, which makes fuel more susceptible to pellet–cladding mechanical interaction induced failure. Therefore, pore pressure calculation becomes vital when evaluating the fuel failure. In this paper pore pressure is calculated as a function of pellet radial local burnup based on the basic characteristics of HBS using Ronchi's correlation. The results indicate that pore pressure will approach a stable value as HBS is developing. A best approximated C value of 55 N/m is recommended for pore pressure calculation

  2. Burnup performances of boron nitride and boron coated nuclear fuels

    International Nuclear Information System (INIS)

    The nuclear fuels of urania (UOV) and 5% and 10% gadolinia (Gd2O3) containing UO2 previously produced by sol-gel technique were coated with first boron nitride (BN) then boron (B) thin layer by chemical vapor deposition (CVD) and also by plasma enhanced chemical vapor deposition (PECVD) techniques to increase the fuel cycle length and to improve the physical properties. From the cross-sectional view of BN and B layers taken from scanning electron microscope (SEM), the excellent adherence of BN onto fuel and B onto BN layer was observed in both cases. The behavior of fuel burnup, depletion of BN and B, the effect of coating thickness and also Gd2O3 content on the burnup performances of the fuels were identified by using the code WIMS-D/4 for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) cores. The optimum thickness ratio of B to BN was found as 4 and their thicknesses were chosen as 40 mm and 10 mm respectively in both reactor types to get extended cycle length. The assemblies consisting of fuels with 5% Gd2O3 and also coated with 10 mm BN and 40 mm B layers were determined as candidates for getting higher burnup in both types of reactors

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

    International Nuclear Information System (INIS)

    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. New burnup calculation of TRIGA IPR-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    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)

  5. Practical issues with implementation of burnup credit in the USA for storage and transportation

    International Nuclear Information System (INIS)

    The US NRC issued an interim staff guidance (ISG8 rev1) allowing for burnup credit applications for storage and transport casks in July of 1999. In over two and a half years there has still not been a license submittal using burnup credit. ISG8 rev1 does not provide sufficient burnup credit to allow loading of 5 wt% enriched fuel in a 32 PWR assembly cask without the addition of absorber rod inserts. Pressure to allow all assemblies to contain inserts from the utility, force continued investigation into alternative levels of burnup credit. Utilities do not wish to measure to confirm burnup. This measurement costs, which range form $10 000 to $50 000 per cask and must be done prior to loading. Since burnup credit is actually only needed for transport, and transport is not expected for many years, many utilities are considering keeping the money in the bank until the time of transport. In order to address the need perceived for additional burnup credit beyond actinide-only burnup credit (ISG8), investigations have moved beyond into assuming moderator exclusion during transport and the use of burnup credit to cover a beyond design basis accident assumption of flooding. Burnup credit analysis requirements for a beyond design basis accident should be less than that for criticality control for normal operation. It is proposed that burnup credit analysis to cover the beyond design basis accident of flooding should be consistent with the beyond design basis dilution event in PWR spent fuel pools. The US NRC precedence for this type of burnup credit allows for all isotopes, a 5% reduction in the delta k of burnup, and an allowable keff of less than 1.0 after biases and uncertainties. (author)

  6. Validation of BGCore System for Burnup Calculations

    International Nuclear Information System (INIS)

    BGCore is a software package for comprehensive computer simulation of nuclear reactor systems and their fuel cycles. BGCore interfaces the Monte Carlo particles transport code MCNP4C with a SARAF module - an independently developed code for calculating fuel composition during irradiation and spent fuel emissions following discharge. In BGCore system, depletion coupling methodology is based on the multi-group approach that significantly reduces computation time and allows tracking of large number of nuclides during calculations. The objective of this study is validation of the BGCore system against well established and verified, state of the art computer codes for thermal and fast spectrum lattices

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

    International Nuclear Information System (INIS)

    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)

  8. Behaviour of Bruce NGS A fuel irradiated to a burnup of ∼ 500 MWh/kgU

    International Nuclear Information System (INIS)

    Fuel elements from three fuel bundles discharged from Bruce NGS A at burnups of 478 to 518 MWh/kgU were examined in the hot cell facilities at Chalk River Laboratories. The examination results show that the fuel operated at higher than expected temperatures, resulting in fission gas releases of up to (and probably exceeding) 24%. Sheathing strains of up to 3% were also observed, indicating a high level of pellet-clad interaction (PCI). This unexpected behaviour is believed to be related to a burnup-dependent reduction in UO2 thermal conductivity. All of the outer elements examined from two of the bundles apparently experienced stress corrosion cracking-related failures in the end cap weld region and sheath as a result of PCI. No failures were observed in the elements examined from the third bundle. The factors believed to have primarily influenced performance are end cap weld geometry and UO2 density. Pellet geometry, stack length and sheath properties may also have influenced performance. The results of these examinations may help to facilitate performance enhancements to existing and future fuel bundle designs. Until such an optimization takes place, the current bundle average burnup limit in Ontario Hydro reactors of 450 MWh/kgU is justified. (Author) (13 refs., 10 figs., tab.)

  9. Self-shielding effects in burnup of Gd used as burnable absorber. Previous studies on its experimental verification

    International Nuclear Information System (INIS)

    Continuing with the domestic 'Burnable Absorbers Research Plan' studies were done to estimate self-shielding effects during Gd2O3 burnup as burnable absorber included in fuel pins of a CAREM geometry. In this way, its burnup was calculated without and with self-shielding. For the second case, were obtained values depending on internal pin radius and the effective one for the homogenized pin. For Gd 157, the burnup corresponding to the first case resulted 52.6 % and of 1.23 % for the effective one. That shows the magnitude of the effects under study. Considering that is necessary to perform one experimental verification, also are presented calculational results for the case to irradiate a pellet containing UO2 (natural) and 8 wt % of Gd2O3, as a function of cooling time, that include: measurable isotopes concentrations, expected activities, and photon spectra for conditions able to be compared with bidimensional calculations with self-shielding. The irradiation time was supposed 30 dpp using RA-3 reactor at 10 MW. (author)

  10. Evaluation of the elastic constants of the high burnup nuclear fuel (U1-y,Gdy)O2

    International Nuclear Information System (INIS)

    Nuclear fuels with burnable poisons are applications in the scope of the nuclear energy for electricity generation. The discharge burnups of this type of fuel are higher than the discharge burnups of pure UO2 fuels with same enrichment of the fissile element. In conditions of high burnups and moderate transients, the fuel rods can be subject to the pellet/cladding mechanical interactions (PCMI). Elastic constants of the fuel and cladding are used in the thermo-mechanical evaluation of the PCMI. The fuel elastic constants are usually expressed as a function of temperature, composition and porosity. Data survey and modeling of elastic constants for the (U1-y,Gdy)O2 fuel are developed in this paper. Due to small amount of the available measurements for the (U1-y,Gdy)O2, the modeling taken into account the data from UO2 and Gd2O3 measurements. Available measurements of the Al2O3 were also used in the model validation. The elastic constants of these materials have been fitted by an analytic equation, which depicts the whole range the dependences. (author)

  11. Results of post-irradiation examination to validate WWER-440 and WWER-1000 fuel efficiency at high burnups

    International Nuclear Information System (INIS)

    During the last 10 years on the basis of commercial operation of WWER reactors, a conversion from three to four year fuel cycle operation has been succeeded for WWER-440 fuel. This paper presents the examinations of fuel rods and fuel assemblies operated at different NPPs of Russia and Eastern Europe. Three WWER-440 fuel assemblies with different burnups and different irradiation in the core (3, 4 and 5 years fuel cycle) have passed full-scale examinations including both: destructive and non-destructive methods. The results of examinations have revealed that the irregularity of the field of the energy release may result in increased cladding oxidation. A validation of WWER-440 and WWER-1000 fuel efficiency during 4 and 5 years fuel cycles is also made on the basis of assessment of fuel rod and assembly mechanical state and changes in their geometry. The status of the fuel column including grain size, fuel swelling, rim layer and fission gas release depending on fuel burnup are investigated. During examinations mechanical properties and oxidation of the cladding, mechanical and corrosion state of the spacer grid, ultimate stress, hardness and plasticity of central tube and relaxation of spring unit are studied at the maximal fuel burnup 64 MWd/kgU for WWER-440 and 58 MWd/kgU for WWER-1000. Based on the examination results for the principal parameters determined fuel resource (variation in form, material structure and properties, corrosion resistance of the claddings, FGR form the fuel, fuel cladding interaction degree) reliable fuel operation at the burnups corresponding to four and five fuel cycles may be predicted. None of fuel efficiency factors in up-to date FA design are limited for operation during five fuel cycles

  12. Corrections and additions to the proposal of a benchmark for core burnup calculations for a WWER-1000 reactor

    International Nuclear Information System (INIS)

    At the nineteenth AER symposium a benchmark on core burnup calculations for WWER-1000 reactors was proposed for further validation and verification of the reactor physics code systems. The work was continued in the framework of a project supported by the German BMU3). During the preparation of the calculations results corrections, refinement and additions the benchmark specification were done. The benchmark includes two stages: the first step comprises the data library preparation for all fuel assembly types used in the core loadings. The second step consists of the 3D core burnup calculation together with calculations of critical states for hot zero power conditions. The benchmark specification contains the description of the fuel assemblies (FA) for the few group data preparation, the core loading patterns and the load follow as well as a set of reference data such as boron acid concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions for successive cycles of a WWER-1000 reactor core. Different reactor physics codes were used to produce solutions. FA burnup codes such as NESSEL, CASMO or HELIOS were used for data preparation. The core calculations were performed using codes such as DYN3D, TRAPEZ as well as several data libraries. The results of the calculations made by different organisations (IBBS, FZD, SSTC) are presented and discussed. The data needed to produce solutions as well as most of the calculated data are attached in the appendices of the paper presented. (Authors)

  13. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    International Nuclear Information System (INIS)

    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 (keff) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of keff. 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 keff values calculated by the participants from the mean value is almost within the band of ±1%Δk/k. The deviations from the averaged calculated fission rate profiles are found to be within ±5% for most cases. (author)

  14. Temperature dependence of 13C 1H one-bond coupling constants of methyl groups in plastic crystals

    Science.gov (United States)

    Aksnes, Dagfinn W.; Balevicius, Vytautas J.; Kimtys, Liudvikas L.

    The temperature dependence of the one-bond 13C 1H coupling constant of the methyl groups in pivalic acid, tert-butyl chloride and hexamethylethane has been studied in the liquid and plastic crystalline phases. A steady decrease in the coupling constant with falling temperature in the plastic crystalline phase has been observed for these organic solids. A maximum change in the 13C 1H coupling constant of 25 Hz has been found after deduction of the effect of overlap of the broadened lines in the methyl quartet. The CNDO/2 calculations indicate that the temperature dependence of the coupling constant is not caused by intramolecular transitions. The significant reduction of the 13C 1H coupling constant is largely attributed to intramolecular dipole-dipole interactions due to a slight anisotropic tumbling of the molecules in the plastic phase.

  15. Time-dependent Multi-group Multidimensional Relativistic Radiative Transfer Code Based On Spherical Harmonic Discrete Ordinate Method

    CERN Document Server

    Tominaga, Nozomu; Blinnikov, Sergei I

    2015-01-01

    We develop a time-dependent multi-group multidimensional relativistic radiative transfer code, which is required to numerically investigate radiation from relativistic fluids involved in, e.g., gamma-ray bursts and active galactic nuclei. The code is based on the spherical harmonic discrete ordinate method (SHDOM) that evaluates a source function including anisotropic scattering in spherical harmonics and implicitly solves the static radiative transfer equation with a ray tracing in discrete ordinates. We implement treatments of time dependence, multi-frequency bins, Lorentz transformation, and elastic Thomson and inelastic Compton scattering to the publicly available SHDOM code. Our code adopts a mixed frame approach; the source function is evaluated in the comoving frame whereas the radiative transfer equation is solved in the laboratory frame. This implementation is validated with various test problems and comparisons with results of a relativistic Monte Carlo code. These validations confirm that the code ...

  16. Dependence of light pulse propagation on its temporal width: Transition from group velocity to c-propagation

    Science.gov (United States)

    Ignesti, Emilio; Tommasi, Federico; Fini, Lorenzo; Cavalieri, Stefano

    2016-07-01

    We show how the velocity of an optical pulse propagating through a dispersive medium depends on the pulse duration. A transition from the group velocity for long pulses to the in-vacuum velocity for short pulses is shown both in experimental results and in theoretical predictions. The temporal duration of the experimental pulses are 150 ps and 3.5 ns. A description of the pulse propagation in terms of the time "center of mass" of the energy flow allows an intuitive overview of the results.

  17. Burnup analysis of the power reactor, 1

    International Nuclear Information System (INIS)

    Several years of endeavors has been devoted to development of three-dimensional nuclear-thermal-hydro-dynamic simulators and research by basing the progress on the merits and demerits of the variational method, the functional approximation method, etc. As the result, the three-dimensional nuclear-thermal-hydro-dynamic code FLORA has been prepared. It has the following features. (1) The executive time is one third -- half as much as that by the convensional programs. (2) Numerical error is small when neutron spectrum mismatches. (3) In the fuels in which the distributions of Gd2O3 and enrichments are localized axially in the reactor core, three-dimensional nuclear-thermal-hydro-dynamic calculations are possible. (4) The transport kernel can be obtained by the coarse mesh method and the functional approximation method. (5) Albedo can be calculated by the two-group diffusion theory. (6) Power distribution can be obtained in the case of partial control rods inserted in the core. The course taken to the preparation, the theoretical background and example calculations with FLORA are described. The present report can be also used as a manual. (auth.)

  18. 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.

  19. Research on Integrity of High Burnup Spent Fuel Under the Long Term Dry Storage

    International Nuclear Information System (INIS)

    Objectives were to acquire the following behaviour data by dynamic load impact tests on high burnup spent fuel rods of BWR and PWR and to improve the guidance of regulation of spent fuel storage and transportation. (1) The limit of load and strain for high burnup fuel in the cask drop accident. (2) The amount of deformation of high burnup fuel rods under dynamic load impact. (3) The amount of fuel pellet material released from fuel rods under dynamic load impact

  20. Nuclear fuel burn-up credit for criticality safety justification of spent nuclear fuel storage systems

    International Nuclear Information System (INIS)

    Burn-up credit analysis of RBMK-1000 an WWER-1000 spent nuclear fuel accounting only for actinides is carried out and a method is proposed for actinide burn-up credit. Two burn-up credit approaches are analyzed, which consider a system without and with the distribution of isotopes along the height of the fuel assembly. Calculations are performed using SCALE and MCNP computer codes

  1. Nuclear fuel behaviour modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    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

  2. Study on burn-up credit and minor actinide in post-irradiation analysis

    International Nuclear Information System (INIS)

    Accuracy of burnup calculation for actinide is very important as to the study of burn-up credit. For minor-actinides such as Am243 and Cm244, however, typical burnup calculation codes are not accurate enough. The accuracy for both nuclides was studied by using the SWAT code. The study showed that the C/E values of both nuclides could be improved at the same time by changing the cross section of Pu242. A study of burnup calculation related to the cross section of Pu242 should be performed to improve the accuracy for both nuclides. (author)

  3. Results of the isotopic concentrations of VVER calculational burnup credit benchmark No. 2(CB2)

    International Nuclear Information System (INIS)

    Results of the nuclide concentrations are presented of VVER Burnup Credit Benchmark No. 2(CB2) that were performed in The Nuclear Technology Center of Cuba with available codes and libraries. The CB2 benchmark specification as the second phase of the VVER burnup credit benchmark is summarized. The CB2 benchmark focused on VVER burnup credit study proposed on the 97' AER Symposium. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and cooling time. The depletion point 'ORIGEN2' code and other codes were used for the calculation of the spent fuel concentration. (author)

  4. Effect of fuel burnup history on neutronic characteristics of WWER-1000 core

    International Nuclear Information System (INIS)

    The paper analyzes fuel burnup history effect on neutronic characteristics of WWER-1000 core with use of the DYN3D codes. The DYN3D code employs the local Pu-239 concentration as an indicator of burnup spectral history. The calculations have been performed for the first four fuel loadings of Khmelnitsky NPP unit 2 and stationary fuel loading with TVSA. The effect of fuel burnup history is shown both on macro-characteristics on the reactor core and on local values of burnup and power

  5. Development of one-energy group, two-dimensional, frequency dependent detector adjoint function based on the nodal method

    International Nuclear Information System (INIS)

    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

  6. LOLA-SYSTEM, JEN-UPM PWR Fuel Management System Burnup Code System

    International Nuclear Information System (INIS)

    1 - Description of program or function: The LOLA-SYSTEM is a part of the JEN-UPM code package for PWR fuel management, scope or design calculations. It is a code package for core burnup calculations using nodal theory based on a FLARE type code. The LOLA-SYSTEM includes four modules: the first one (MELON-3) generates the constants of the K-inf and M2 correlations to be input into SIMULA-3. It needs the K-inf and M2 fuel assembly values at different conditions of moderator temperature, Boron concentration, burnup, etc., which are provided by MARIA fuel assembly calculations. The main module (SIMULA-3) is the core burnup calculation code in three dimensions and one group of energy. It normally uses a geometrical representation of one node per fuel assembly or per quarter of fuel assembly. It has included a thermal hydraulic feedback on flow and voids and criticality searches on boron concentration and control rods insertion. The CONCON code makes the calculation of the albedo, transport factors, K-inf and M2 correction factors to be input into SIMULA-3. The calculation is made in the XY transversal plane. The CONAXI code is similar to CONCON, but in the axial direction. 2 - Method of solution: MELON-3 makes a mean squares fit of K-inf and M2 values at different conditions in order to determine the constants of the feedback correlations. SIMULA-3 uses a modified one-group nodal theory, with a new transport kernel that provides the same node interface leakages as a fine mesh diffusion calculation. CONCON and CONAXI determine the transport and correction factors, as well as the albedo, to be input into SIMULA-3. They are determined by a method of leakages equivalent to the detailed diffusion calculation of CARMEN or VENTURE; these factors also include the heterogeneity effects inside the node. 3 - Restrictions on the complexity of the problem: Number of axial nodes less than or equal 34. Number of material types less than or equal 30. Number of fuel assembly types less

  7. 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)

    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)

  8. The high burn-up restructuring of fuel, the so called rim effect: a consideration of its impact on steady-state and transient behaviour

    International Nuclear Information System (INIS)

    The report sets out to investigate our current understanding on the occurrence, properties and effect of restructured fuel material as observed in the pellet rim of high burn-up fuel. It appears that restructuring occurs solely as a function of burn-up and temperature. In this case, the driving force is likely to be the accumulation of irradiation damage and fission products. There are differing theories for which of these dominate the transformation, although the conditions for HBS formation are well established. The major part of the report addressed the influence this HBS has on fuel performance and to this end a simple model for the restructuring thickness and local swelling was inserted into a code to calculate fuel temperatures and Fission Gas Release. By running this code for three Halden experiments it was shown that inclusion of a rim had a measurable effect on both centreline temperatures and FGR. The calculations have shown that the effect of the HBS increases FGR by around 30% or 10% depending on the assumption made as to the HBS matrix thermal conductivity. From an appraisal of several experiments, it is proposed that fuel immediately prior to restructuring has the most influence on PCMI and fuel dispersal in high burn-up RIA. It is suggested that in the forthcoming IFA-655 test, the transient behaviours of fuel at less than the maximum burn-up of 100 MWd/kg is investigated. (Author)

  9. Destructive radiochemical analysis of uraniumsilicide fuel for burnup determination

    Energy Technology Data Exchange (ETDEWEB)

    Gysemans, M.; Bocxstaele, M. van; Bree, P. van; Vandevelde, L.; Koonen, E.; Sannen, L. [SCK-CEN, Boeretang, Mol (Belgium); Guigon, B. [CEA, Centre de Cadarache, Saint Paul lez Durance (France)

    2004-07-01

    During the design phase of the French research reactor Jules Horowitz (RJH) several types of low enriched uranium fuels (LEU), i.e. <20% {sup 235}U enrichment, are studied as possible candidate fuel elements for the reactor core. One of the LEU fuels that is taken into consideration is an uraniumsilicide based fuel with U{sub 3}Si{sub 2} dispersed in an aluminium matrix. The development and evaluation of such a new fuel for a research reactor requires an extensive testing and qualification program, which includes destructive radiochemical analysis to determine the burnup of irradiated fuel with a high accuracy. In radiochemistry burnup is expressed as atom percent burnup and is a measure for the number of fissions that have occurred per initial 100 heavy element atoms (%FIMA). It is determined by measuring the number of heavy element atoms in the fuel and the number of atoms of selected key fission products that are proportional to the number of fissions that occurred during irradiation. From the few fission products that are suitable as fission product monitor, the stable Nd-isotopes {sup 143}Nd, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148Nd}, {sup 150}Nd and the gamma-emitters {sup 137}Cs and {sup 144}Ce are selected for analysis. Samples form two curved U{sub 3}Si{sub 2} plates, with a fuel core density of 5.1 and 6.1 g U/cm{sup 3} (35% {sup 235}U) and being irradiated in the BR2 reactor of SCK x CEN{sup [1]}, were analyzed. (orig.)

  10. Simulation of fuel cycles with minor actinide management using a fast burnup calculation tool

    International Nuclear Information System (INIS)

    The paper presents a fast and flexible burnup model for fuel cycle simulations which is based on the description of the one-group cross-sections as analytic functions of the isotopic composition. This was accomplished by multi-dimensional regression based on the results of numerous core calculations. The developed model is able to determine the spent fuel composition in reasonable CPU time, and was integrated into a simplified fuel cycle model containing Gas Cooled Fast Reactors (GFR) and conventional light water reactors (LWRs). The fuel cycle simulations revealed an advantageous effect of increased minor actinide content in the GFR core on the fuel utilization parameters. In order to explore the processes that lay behind this effect the neutronics balance of the GFR was investigated in equilibrium cycle conditions. (author)

  11. Framatome-ANP extended burnup experience and views on LWR fuels

    International Nuclear Information System (INIS)

    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)

  12. Nondestructive analysis of RA reactor fuel burnup, Program for burnup calculation base on relative yield of 106Ru, 134Cs and 137Cs in the irradiated fuel

    International Nuclear Information System (INIS)

    Burnup of low enriched metal uranium fuel of the RA reactor is described by two chain reactions. Energy balance and material changes in the fuel are described by systems of differential equations. Numerical integration of these equations is base on the the reactor operation data. Neutron flux and percent of Uranium-235 or more frequently yield of epithermal neutrons in the neutron flux, is determined by iteration from the measured contents of 106Ru, 134Cs and 137Cs in the irradiated fuel. The computer program was written in FORTRAN-IV. Burnup is calculated by using the measured activities of fission products. Burnup results are absolute values

  13. Modelling of fission gas behaviour in high burnup nuclear fuel

    International Nuclear Information System (INIS)

    The safe and economic operation of nuclear power plants (NPPs) requires that the behaviour and performance of the fuel can be calculated reliably over its expected lifetime. This requires highly developed codes that treat the nuclear fuel in a general manner and which take into account the large number of influences on fuel behaviour, in particular the trend of NPP operators to increase the fuel burnup. With higher burnup, more fission events impact the material characteristics of the fuel and significant restructuring can be observed. At local burnups in excess of 60-75 MWd/kgU, the microstructure of nuclear fuel pellets differs markedly from the as-fabricated structure. This high burnup structure (HBS) is characterised by three principal features: 1) low matrix xenon concentration, 2) sub-micron grains and 3) a high volume fraction of micrometer-sized pores. The peculiar features of the HBS affect the fuel performance and safety; the large retention of fission gas within the HBS could lead to significant gas release at high burnups, either through the degradation of thermal conductivity or through direct release. The present work has focussed on the development and evaluation of HBS fission gas transport models, especially on two features: the equilibrium xenon concentration in the matrix of the HBS in UO2 fuel pellets, and the growth of the HBS porosity and its effect on fission gas release. A steady-state fission gas model has been developed to examine the importance of grain boundary diffusion for the gas dynamics in the HBS. It was possible to simulate the ∼0.2 wt% experimentally observed xenon concentration. The value of the grain boundary diffusion coefficient is not important for diffusion coefficient ratios in excess of ∼10”4. The model exhibits a high sensitivity to principally three parameters: the grain diffusion coefficient, the bubble number density and the re-solution rate coefficient. The model can reproduce the observed HBS xenon depletion

  14. The Design Method for the ATR High Burnup MOX Fuel

    International Nuclear Information System (INIS)

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has developed the advanced thermal reactor (ATR). PNC is demonstrating MOX fuel utilization in a prototype of ATR, Fugen (165 MWe), in which 638 MOX fuel assemblies have been loaded without a failure since 1979. PNC is developing the high burn-up MOX fuel for the ATR to contribute to MOX fuels for thermal reactors. The statistical design evaluation method that included the MOX fuel rod performance evaluation code 'FEMAXI-ATR' was developed for the ATR high bum-up MOX fuel rod; it was verified that the integrity of the fuel could be maintained over the whole irradiation period

  15. Development and verification of Monte Carlo burnup calculation system

    International Nuclear Information System (INIS)

    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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide [Japan Atomic Research Institute, Toaki, Ibaraki (Japan)] [and others

    1996-03-01

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

  17. Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

    International Nuclear Information System (INIS)

    The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup

  18. Overview of the burnup credit activities of the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA)

    International Nuclear Information System (INIS)

    This article summarises activities of the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) Expert Group on Burnup Credit Criticality, a subordinate group to the Working Part 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 group 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 Expert Group on Burnup Credit Criticality is to demonstrate that the available criticality safety calculational tools are appropriate for application to irradiated (burned) nuclear fuel systems and that a reasonable safety margin can be established. The method established by the expert group 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 is being addressed by the expert group. 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 group 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)

  19. Chemical form of fission products in high burnup fuels

    International Nuclear Information System (INIS)

    In order to make a proper assessment of candidate materials for advanced high-burnup fuels, thermochemical studies of fuel materials have been performed. Using data from the ECN thermochemical database (TBASE), which has been updated and extended for the present work, the suitability of various advanced fuel materials and inert matrices is studied. Detailed thermodynamic equilibrium calculations are performed for Pu0.42U0.58O2 and Pu0.40U0.60N for values of the burnup up to 200 MWd/kgHM. The formation of metallic phases, the pressure buildup and the stability of nitride or oxide phases is studied for each fuel type. The results for the chemical form of the solid fission products are given. The chemical aspects of the use of the inert matrix spinel (MgAl2O4) in combination with oxide fuel will be discussed. Experimental research on the compatibility of various types of inert matrices (nitrides, spinel) is in progress at ECN. (author)

  20. Fission-product burn-up in fast reactors

    International Nuclear Information System (INIS)

    In fast reactors where breeding is emphasized the burn-up of fission products can be of considerable importance. Statistical estimates of fission-product cross-sections are combined with recent yield data for the various fissionable species to estimate the gross fission-product cross-section as a function of irradiation time in a number of fast reactor spectra with various fuels. Because of gaps in yield data for some of the fuel species, it is necessary to interpolate on the yield curves in some cases. The chain yield for a given mass is then apportioned among the chain members through use of the equal charge displacement recipe. The cross-sections estimated for U235 fission products by previous authors are supplemented by estimates for fission products important for other fuels. A range of such spectra is considered. These spectra are characterized by the index (average (Ε-1/2)) in the spectra. The sensitivity of the gross poisoning and its burn-up with respect to spectrum variations are considered. The results are also expressed in terms of a few pseudo-fission products, so that changes in effective cross-section of fission products with irradiation can be taken into account in a simple computational fashion. (author)

  1. Development of advanced cladding material for burnup extension

    International Nuclear Information System (INIS)

    The development of new cladding materials is one of the critical issues on burnup extension. The practical life of Zircaloy would be limited by the growth of oxide films and by the ductility loss due to hydride precipitation, oxygen absorption and radiation damage. In the case of high burnup using MOX fuels, the low neutron adsorption cross section of Zircaloy is not a dominant factor for selecting the cladding material, because MOX fuels can be enriched up to 20%Pu. Austenitic stainless steel, titanium alloy, niobium alloy, ferritic steel and nickel base superalloy are considered as candidate materials. The corrosion resistance, mechanical properties and the irradiation resistance of these materials were examined for evaluating the practical possibility as a cladding material. The austenitic stainless steel with high g phase stability was selected as the primary candidate material. However, it is required to improve the resistance to irradiation associated stress corrosion cracking through the experience in LWR plants. In the JAERI, the austenitic stainless steel with intergranular corrosion resistance has been developed by the adjustment of the chemical composition, the modification of the metallographic structure by thermo-mechanical treatment and the purification by electron beam melting. (author)

  2. High Burnup UO2 Fuel Pellets with Dopants for WWER

    International Nuclear Information System (INIS)

    The currently achieved level of design and technology developments provided for the implementation of the fuel cycle (4x1) in WWER at the maximal design burnup of 56 MW.day/kgU per FA. Presently in Russia the program is under way to improve the technical and economic parameters of WWER fuel cycles characterized by an increased fuel usability. To meet the requirements placed on the new fuel that ensures the reliable operation under conditions of higher burnups complex activities are under way to optimize the composition and microstructure of fuel pellets as applied to WWER. This paper describes a general approach to providing the stimulated composition and microstructure of fuel via introducing various dopants. Aside from this, the paper presents the experimentally results of studies into the main technologic and operational characteristics of dopant containing fuel pellets including higher grain sizes, pores distribution and oxygen to metal ratio. The results of the experiments made it possible to work out the pilot commercial process of the modified fuel fabrication, to manufacture pellet batches to be semi-commercially operated at NPP with WWER. (author)

  3. Bacterial histo-blood group antigens contributing to genotype-dependent removal of human noroviruses with a microfiltration membrane.

    Science.gov (United States)

    Amarasiri, Mohan; Hashiba, Satoshi; Miura, Takayuki; Nakagomi, Toyoko; Nakagomi, Osamu; Ishii, Satoshi; Okabe, Satoshi; Sano, Daisuke

    2016-05-15

    We demonstrated the genotype-dependent removal of human norovirus particles with a microfiltration (MF) membrane in the presence of bacteria bearing histo-blood group antigens (HBGAs). Three genotypes (GII.3, GII.4, and GII.6) of norovirus-like particles (NoVLPs) were mixed with three bacterial strains (Enterobacter sp. SENG-6, Escherichia coli O86:K61:B7, and Staphylococcus epidermidis), respectively, and the mixture was filtered with an MF membrane having a nominal pore size of 0.45 μm. All NoVLP genotypes were rejected by the MF membrane in the presence of Enterobacter sp. SENG-6, which excreted HBGAs as extracellular polymeric substances (EPS). This MF membrane removal of NoVLPs was not significant when EPS was removed from cells of Enterobacter sp. SENG-6. GII.6 NoVLP was not rejected with the MF membrane in the presence of E. coli O86:K61:B7, but the removal of EPS of E. coli O86:K61:B7 increased the removal efficiency due to the interaction of NoVLPs with the exposed B-antigen in lipopolysaccharide (LPS) of E. coli O86:K61:B7. No MF membrane removal of all three genotypes was observed when S. epidermidis, an HBGA-negative strain, was mixed with NoVLPs. These results demonstrate that the location of HBGAs on bacterial cells is an important factor in determining the genotype-dependent removal efficiency of norovirus particles with the MF membrane. The presence of HBGAs in mixed liquor suspended solids from a membrane bioreactor (MBR) pilot plant was confirmed by immune-transmission electron microscopy, which implies that bacterial HBGAs can contribute to the genotype-dependent removal of human noroviruses with MBR using MF membrane. PMID:27095709

  4. Friend or ally: whether cross-group contact undermines collective action depends on what advantaged group members say (or don't say).

    Science.gov (United States)

    Becker, Julia C; Wright, Stephen C; Lubensky, Micah E; Zhou, Shelly

    2013-04-01

    Previous research shows that positive contact with members of advantaged groups can undermine collective action among the disadvantaged. The present work provides the first experimental evidence of this effect and introduces a moderator which highlights the fundamental role of communication about perceptions of the legitimacy of intergroup inequality. Study 1 (N = 267) focused on the lesbian/gay/bisexual/transgendered community's struggle for same-sex marriage in California. In Study 2 (N = 81), cross-group contact was initiated between members of two universities that differ in social status. Results revealed that positive cross-group contact undermined public collective action among the disadvantaged when the advantaged-group partner described their group's advantaged position as legitimate or when they did not communicate their feelings about intergroup inequality (leaving them ambiguous). In contrast, when the advantaged-group partner clearly described the intergroup inequality as illegitimate, cross-group contact did not undermine participation in public collective action. PMID:23504760

  5. Methods used in burn-up determination of the irradiated fuel rods at TRIGA reactor

    International Nuclear Information System (INIS)

    A short presentation of the methods used at INR TRIGA reactor for the burn-up determination is given together with some considerations on ORIGEN 2 computer code used for calculating fission products activities and nuclide concentration. Burn-up is determined by gamma spectroscopy and thermal power monitoring. (Author)

  6. Determination of axial profit performed burnup credit by SCALE 4.3-system

    International Nuclear Information System (INIS)

    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

  7. An overview of burnup credit application in spent nuclear fuel management

    International Nuclear Information System (INIS)

    The current status of burnup credit application has been overviewed for spent nuclear fuel management. It was revealed that the use of burnup credit is practically limited to spent nuclear fuel storage, for which selected actinides-only are taken into account

  8. Specific behaviour aspects at extended burnup operation of PHWR nuclear fuels

    International Nuclear Information System (INIS)

    In order to evaluate the influence of burnup extension on PHWR nuclear fuel performance, the paper presents and discusses the specific potentially life-limiting factors at extended burnup for this type of fuel using recent experimental evidence and making a direct comparison with LWR fuel performance. (Author)

  9. Burnup calculations using the ORIGEN code in the CONKEMO computing system

    International Nuclear Information System (INIS)

    This article describes the CONKEMO computing system for kinetic multigroup calculations of nuclear reactors and their physical characteristics during burnup. The ORIGEN burnup calculation code has been added to the system. The results of an international benchmark calculation are also presented. (author)

  10. Microstructural characterization of high burn-up mixed oxide fast reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Teague, Melissa, E-mail: melissa.teague@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Gorman, Brian; King, Jeffrey [Colorado School of Mines, 1500 Illinois St, Golden, CO 80401 (United States); Porter, Douglas; Hayes, Steven [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2013-10-15

    High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column were observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.

  11. PWR rod ejection accident: uncertainty analysis on a high burn-up core configuration

    Energy Technology Data Exchange (ETDEWEB)

    Le Pallec, J.C.; Studer, E.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Service d' Etudes de Reacteurs et de Modelisation Avancee (DEN/SERMA), 91 - Gif sur Yvette (France)

    2003-07-01

    With the increasing of the discharge burn-up assembly, the rod ejection accident (REA) methodology based on the analyse of the hot spot from a decoupling methods of calculation does not allow to ensure the respect of safety criteria. The main reason is that the irradiated fuel certainly less solicited thermally is in the other hand more sensitive to a transient due to a rod ejection. Thus, the hot spot is not necessarily the sensitive point of the core. In the framework of high burn-up configurations, a new methodology tends to replace the former. It characterizes by the use of a best-estimate 3-dimensional modelling: coupling of the thermal hydraulics and neutronics, taking in account fuel properties depending on irradiation. To ensure the conservatism of the modelling response, this new approach has to be followed by an uncertainties analysis. Inputs from the benchmark RIA TMI-1 conducted by IRSN (France), NRC (United State of America) and KI (Russian) are used to perform a first analysis. The response of the modelling is the enthalpy deposited in an assembly. The analysis is based on the Design of Experiments (DoE) that permits to measure the weight of the main parameters and their interactions on the response. These last cannot be disregarded because they represent up to 20% of the penalizing uncertainty. This study shows that the main fuel modifications due to irradiation (radial power distribution, thermal properties degradation) have to be taken into account in a realistic thermal modelling during a strong transient.

  12. Important fission product nuclides identification method for simplified burnup chain construction

    International Nuclear Information System (INIS)

    A method of identifying important fission product (FP) nuclides which are included in a simplified burnup chain is proposed. This method utilizes adjoint nuclide number densities and contribution functions which quantify the importance of nuclide number densities to the target nuclear characteristics: number densities of specific nuclides after burnup. Numerical tests with light water reactor (LWR) fuel pin-cell problems reveal that this method successfully identifies important FP nuclides included in a simplified burnup chain, with which number densities of target nuclides after burnup are well reproduced. A simplified burnup chain consisting of 138 FP nuclides is constructed using this method, and its good performance for predictions of number densities of target nuclides and reactivity is demonstrated against LWR pin-cell problems and multi-cell problem including gadolinium-bearing fuel rod. (author)

  13. A survey of previous and current industry-wide efforts regarding burnup credit

    International Nuclear Information System (INIS)

    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

  14. Analysis of burnup credit on spent fuel transport / storage casks - estimation of reactivity bias

    International Nuclear Information System (INIS)

    Chemical analyses of high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins were carried out. Measured data of nuclides' composition from U234 to P 242 were used for evaluation of ORIGEN-2/82 code and a nuclear fuel design code (NULIF). Critically calculations were executed for transport and storage casks for 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for axial and horizontal profiles of burnup, and historical void fraction (BWR), operational histories such as control rod insertion history, BPR insertion history and others, and calculational accuracy of ORIGEN-2/82 on nuclides' composition. This study shows that introduction of burnup credit has a large merit in criticality safety analysis of casks, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for the present reactivity bias evaluation and showed the possibility of simplifying the reactivity bias evaluation in burnup credit. (authors)

  15. Fuel Element Designs for Achieving High Burnups in 220 MW(e) Indian PHWRs

    International Nuclear Information System (INIS)

    Presently 19-element natural uranium fuel bundles are used in 220 MW(e) Indian PHWRs. The core average design discharge burnup for these bundles is 7000 MW·d/Te U and maximum burnup for assembly goes upto of 15 000 MWD/Te U. Use of fuel materials like MOX, Thorium, slightly enriched uranium etc in place of natural uranium in 19-element fuel bundles, in 220 MW(e) PHWRs is being investigated to achieve higher burnups. The maximum burnup investigated with these bundles is 30 000 MW·d/Te U. In PHWR fuel elements no plenum space is available and the cladding is of collapsible type. Studies have been carried out for different fuel element target burnups with different alternative concepts. Modification in pellet shape and pellet parameters are considered. These studies for the PHWR fuel elements/assemblies have been elaborated in this paper. (author)

  16. Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

    International Nuclear Information System (INIS)

    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 MTRPC 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, MTRPC Package, Empirical Formula For Fuel Burn-Up.

  17. Burn-up measurements at TRIGA fuel elements containing strong burnable poison

    International Nuclear Information System (INIS)

    The reactivity method of determining the burn-up of research reactor fuel elements is applied to the highly enriched FLIP elements of TRIGA reactors. In contrast to other TRIGA fuel element types, the reactivity of FLIP elements increases with burn-up due to consumption of burnable poison. 33 fuel elements with burn-up values between 3% and 14% were investigated. The experiments showed that variations in the initial fuel composition significantly influence the reactivity and, consequently, increase the inaccuracy of the burn-up measurements. Particularly important are variations in the initial concentration of erbium, which is used as burnable poison in FLIP fuel. A method for reducing the effects of the material composition variations on the measured reactivity is presented. If it is applied, the accuracy of the reactivity method for highly poisoned fuel elements becomes comparable to the accuracy of other methods for burn-up determination. (orig.)

  18. Measurement of gamma attenuation coefficients in UO2 and zirconium for self-absorption corrections of burn-up determination

    International Nuclear Information System (INIS)

    UO2 pellets from ALUOX fuel elements were used in measuring the absorption coefficient of gamma radiation in UO2. 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 UO2 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.)

  19. RAPID program to predict radial power and burnup distribution of UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Song, Jae Sung; Bang, Je Gun; Kim, Dae Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-02-01

    Due to the radial variation of the neutron flux and its energy spectrum inside UO{sub 2} fuel, the fission density and fissile isotope production rates are varied radially in the pellet, and it becomes necessary to know the accurate radial power and burnup variation to predict the high burnup fuel behavior such as rim effects. Therefore, to predict the radial distribution of power, burnup and fissionable nuclide densities in the pellet with the burnup and U-235 enrichment, RAPID(RAdial power and burnup Prediction by following fissile Isotope Distribution in the pellet) program was developed. It considers the specific radial variation of the neutron reaction of the nuclides while the constant radial variation of neutron reaction except neutron absorption of U-238 regardless of the nuclides, the burnup and U-235 enrichment is assumed in TUBRNP model which is recognized as the one of the most reliable models. Therefore, it is expected that RAPID may be more accurate than TUBRNP, specially at high burnup region. RAPID is based upon and validated by the detailed reactor physics code, HELIOS which is one of few codes that can calculates the radial variations of the nuclides inside the pellet. Comparison of RAPID prediction with the measured data of the irradiated fuels showed very good agreement. RAPID can be used to calculate the local variations of the fissionable nuclide concentrations as well as the local power and burnup inside that pellet as a function of the burnup up to 10 w/o U-235 enrichment and 150 MWD/kgU burnup under the LWR environment. (author). 8 refs., 50 figs., 1 tab.

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

    International Nuclear Information System (INIS)

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

  1. Dependency Defence and Dependency Analysis Guidance. Volume 2: Appendix 3-8. How to analyse and protect against dependent failures. Summary report of the Nordic Working Group on Common Cause Failure Analysis

    International Nuclear Information System (INIS)

    The safety systems in Nordic nuclear power plants are characterised by substantial redundancy and/or diversification in safety critical functions, as well as by physical separation of critical safety systems, including their support functions. Viewed together with the evident additional fact, that the single failure criterion has been systematically applied in the design of safety systems, this means that the plant risk profile as calculated in existing PSA:s is usually strongly dominated by failures caused by dependencies resulting in the loss of more than one system sub. The overall objective with the working group is to support safety by studying potential and real CCF events, process statistical data and report conclusions and recommendations that can improve the understanding of these events eventually resulting in increased safety. The result is intended for application in NPP operation, maintenance, inspection and risk assessments. The NAFCS project is part of the activities of the Nordic PSA Group (NPSAG), and is financed jointly by the Nordic utilities and authorities. The work is divided into one quantitative and one qualitative part with the following specific objectives: Qualitative objectives-The goal with the qualitative analysis is to compile experience data and generate insights in terms of relevant failure mechanisms and effective CCF protection measures. The results shall be presented as a guide with checklists and recommendations on how to identify current CCF protection standard and improvement possibilities regarding CCF defences decreasing the CCF vulnerability. Quantitative objectives-The goal with the quantitative analysis is to prepare a Nordic C-book where quantitative insights as Impact Vectors and CCF parameters for different redundancy levels are presented. Uncertainties in CCF data shall be reduced as much as possible. The high redundancy systems sensitivity to CCF events demand a well structured quantitative analysis in support of

  2. Dependency Defence and Dependency Analysis Guidance. Volume 1: Summary and Guidance (Appendix 1-2). How to analyse and protect against dependent failures. Summary report of the Nordic Working group on Common Cause Failure Analysis

    International Nuclear Information System (INIS)

    The safety systems in Nordic nuclear power plants are characterised by substantial redundancy and/or diversification in safety critical functions, as well as by physical separation of critical safety systems, including their support functions. Viewed together with the evident additional fact, that the single failure criterion has been systematically applied in the design of safety systems, this means that the plant risk profile as calculated in existing PSA:s is usually strongly dominated by failures caused by dependencies resulting in the loss of more than one system sub. The overall objective with the working group is to support safety by studying potential and real CCF events, process statistical data and report conclusions and recommendations that can improve the understanding of these events eventually resulting in increased safety. The result is intended for application in NPP operation, maintenance, inspection and risk assessments. The NAFCS project is part of the activities of the Nordic PSA Group (NPSAG), and is financed jointly by the Nordic utilities and authorities. The work is divided into one quantitative and one qualitative part with the following specific objectives: Qualitative objectives-The goal with the qualitative analysis is to compile experience data and generate insights in terms of relevant failure mechanisms and effective CCF protection measures. The results shall be presented as a guide with checklists and recommendations on how to identify current CCF protection standard and improvement possibilities regarding CCF defences decreasing the CCF vulnerability. Quantitative objectives-The goal with the quantitative analysis is to prepare a Nordic C-book where quantitative insights as Impact Vectors and CCF parameters for different redundancy levels are presented. Uncertainties in CCF data shall be reduced as much as possible. The high redundancy systems sensitivity to CCF events demand a well structured quantitative analysis in support of

  3. 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)

    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)

  4. 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).

  5. CDFR design and performance implications of extended burn-up

    International Nuclear Information System (INIS)

    The restrained core was adopted for the UK Commercial Demonstration Fast Reactor (CDFR) in the mid 1970's. The reasons for its adoption are still valid today although increased importance is now given to its seismic performance. During an earthquake of ''safe shutdown'' intensity the rigid restraint barrel, small clearances of the restraint pads and the cross sectional strength of the sub-assemblies, particularly at the upper restraint level adequately maintain the relative positions of sub-assemblies. A general analysis of sub-assembly motions with simplifying but conservative assumptions indicates that the transient reactivity insertions due to horizontal core movements in the safe shutdown earthquakes will be less than 1$. A programme of experiments is being initiated to support the analysis and demonstrate that the present predictions are pessimistic. The effect of higher burn-up on CDFR performance will be discussed in this paper

  6. Introduction of new flasks for high burnup spent fuel

    International Nuclear Information System (INIS)

    New flasks have been designed to transport the high burnup spent fuels now becoming available from the world's nuclear power stations. Two versions have been designed: Excellox 6 for 5 metre PWR fuels and Excellox 7 with increased neutron shielding for 4.5 metre PWR and BWR fuels arising in Japan. The designs of these flasks have been finalised; Excellox 6 has been approved and validated as a Type B(U)F package and the first two have been manufactured and are now in routine service, with a third at an advanced stage of manufacture. The Excellox 7 design is ready for manufacture when service requirements for it have been settled. An account is given of the final adjustments to the design in the course of manufacture, the main steps and tests in the manufacturing process and the commissioning tests at the reprocessing and reactor sites. The entry of the flasks into service is reviewed. (author)

  7. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    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)

  8. 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.

  9. Assessment of reactivity transient experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  10. BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies

    International Nuclear Information System (INIS)

    BURNCAL is a Fortran computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. The code uses output parameters generated by the Monte Carlo neutronics code MCNP to determine the isotopic inventory as a function of time and power density. The code allows for multiple fueled regions to be analyzed. The companion code, RELOAD, can be used to shuffle fueled regions or reload regions with fresh fuel. BURNCAL can be used to study the reactivity effects and isotopic inventory as a function of time for a nuclear reactor system. Neutron transmutation, fission, and radioactive decay are included in the modeling of the production and removal terms for each isotope of interest. For a fueled region, neutron transmutation, fuel depletion, fission-product poisoning, actinide generation, and burnable poison loading and depletion effects are included in the calculation. Fueled and un-fueled regions, such as cladding and moderator, can be analyzed simultaneously. The nuclides analyzed are limited only by the neutron cross section availability in the MCNP cross-section library. BURNCAL is unique in comparison to other burnup codes in that it does not use the calculated neutron flux as input to other computer codes to generate the nuclide mixture for the next time step. Instead, BURNCAL directly uses the neutron absorption tally/reaction information generated by MCNP for each nuclide of interest to determine the nuclide inventory for that region. This allows for the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed

  11. Experimental control of burn-up calculations for high temperature reactor fuel by introduction of a special alpha spectrometric method for the determination of transuranium content. An attempt to establish isotopic correlations

    International Nuclear Information System (INIS)

    In the field of high-temperature-reactor (HTR) fuel investigation there is a great interest in the experimental and calculational determination of heavy metal content under the aspects of burn-up physics and for the prediction of reliable data for reprocessing and waste management. Using a laser-micro-boring preparation method, high resolution alpha-spectroscopy and sophisticated computer decomposition programs we identify qualitatively and quantitatively most of the important actinide isotopes in irradiated HTR-fuel. Additionally we use data, delivered by gamma- and mass-spectroscopy of the same fuel samples. The evaluated results are compared with calculational results from the burn-up code ORIGEN, using a special generated HTR-neutron-cross-section library. In a first step we determine new cross sections for the uranium and plutonium isotopes depending on the irradiation conditions. In a second step we calculate correlations between the heavy metal isotopes and the burn-up or the fission products

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  14. Development and Applications of a Prototypic SCALE Control Module for Automated Burnup Credit Analysis

    International Nuclear Information System (INIS)

    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 keff 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-02-01

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

  16. Radiochemical burnup determination and isotope analysis of four IFA 148 samples: Ris/o/ Fission Gas Project

    Energy Technology Data Exchange (ETDEWEB)

    Mogensen, M.; Larsen, E.; Funck, J.; Strauss, T.R.

    1981-06-01

    In the frame of the Ris/o/ Fission Gas Project, radiochemical burnup determinations (Nd-148) and heavy isotope analyses have been carried out on four IFA 148 samples. The Nd-148 burnup determinations serve as a calibration of the Cs-137 gamma scans so that the absolute burnup can be determined in all axial positions from the Cs-137 curves and the Nd-148 burnup results. The analysis of uranium and transuranium isotopes have the objective of establishing the distribution of the total burnup on U-235, Pu-239 and Pu-241 fissions. 6 refs., 9 figs., 3 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Holly R. Trellue

    1998-12-01

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

  18. Development of burnup analysis system for fast reactor (3) (Contract research)

    International Nuclear Information System (INIS)

    Improvement of the 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 the previous study, considerable improvement of the 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 purposes, it is necessary to improve the prediction accuracy of 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 the prediction accuracy. In the previous study on 'Development of Burnup Analysis System for Fast Reactors (2)' in FY2006, design and implementation of models for detailed geometry of assembly, fuel loading pattern and so on, accompanied with specification and implementation of input file handling to construct data models. In the present study, a prototype system has been implemented in which functionalities are embedded for calculation of macroscopic cross section, core calculation and burnup calculation applying the fruits of the study 'Development of a Framework for the Neutronics Analysis System for Next

  19. Alternatives for implementing burnup credit in the design and operation of spent fuel transport casks

    International Nuclear Information System (INIS)

    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

  20. PIE Results and New Techniques Applied for 55GWd/t High Burnup Fuel of PWR

    International Nuclear Information System (INIS)

    Post-irradiation examinations (PIE) for 55GWd/t high burnup fuel which had been irradiated at a domestic PWR plant was conducted at the fuel hot laboratory of the Nuclear Development Corporation (NDC). In this PIE, such new techniques as the clamping for axial tensile test and the pellets density measurement method for high burnup fuels were used in addition to existing techniques to confirm the integrity of 55GWd/t high burnup fuel. The superiority of improved corrosion-resistant claddings over currently used current Zircaloy-4 claddings in terms of corrosion-resistance was also confirmed. This paper describes the PIE results and the advanced PIE techniques. (author)

  1. 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)

  2. Development of a Burnup Program based on the Krylov Subspace Method

    International Nuclear Information System (INIS)

    The depletion calculation of the DeCART code has been performed by the support of the ORIGEN code. Recently, a burnup program based on the Krylov subspace method is developed and implemented to the DeCART code. Numerical solution for the burnup equation by the Krylov subspace method is well described. Therefore, this paper describes the Krylov subspace method for a burnup equation briefly in Section 2, and focuses on the DeCART solution for a pin cell problem by comparing it with the HELIOS solution

  3. Development of a Burnup Program based on the Krylov Subspace Method

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jin-Young; Shim, Hyung-Jin; Kim, Kang-Seog; Song, Jae-Seung; Lee, Chung-Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    The depletion calculation of the DeCART code has been performed by the support of the ORIGEN code. Recently, a burnup program based on the Krylov subspace method is developed and implemented to the DeCART code. Numerical solution for the burnup equation by the Krylov subspace method is well described. Therefore, this paper describes the Krylov subspace method for a burnup equation briefly in Section 2, and focuses on the DeCART solution for a pin cell problem by comparing it with the HELIOS solution.

  4. Room temperature leaching of labile radioactivity from irradiated PWR fuel according to the burnup

    International Nuclear Information System (INIS)

    Three PWR UO2 spent fuel specimens with average burnup of 22, 37 and 47 GWd tU-1 were submitted to sequential mode leaching in granitic groundwater for 62 cumulative days. The leaching rate decreased versus increasing contact time from 10-3 d-1 to 10-5 d-1. The 90Sr release appeared to be independent of the burnup with rates 2 orders of magnitude lower than for Cs but higher than the U and Pu release rates; both of the latter elements reached saturation rapidly, giving concentration values of 50-800 ppb and 0.1-10 ppb respectively, irrespective of the burnup. (authors)

  5. Pool inspection techniques for surveillance and further development of high burnup fuel assemblies

    International Nuclear Information System (INIS)

    The pool inspection techniques which have been used in fuel assembly surveillance programs for many years are suitable for high burnup fuel assemblies too. The techniques have been adapted to the requirements of new fuel assembly concepts with higher burnup potential. For high burnup, the emphasis within the scope of examination techniques available has shifted towards a characterization of the corrosion behaviour and surveillance of the geometrical dimensions of the fuel assemblies. In order to accomplish these tasks complementary techniques will have to be developed. (orig.)

  6. An analysis of burnup reactivity credit for reactor RA spent fuel storage

    International Nuclear Information System (INIS)

    The need for increasing the spent fuel storage capacity has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the criticality safety analysis methodology used to investigate the sensitivity of storage system reactivities to changes in fuel burnup. Results representing the validation of the methods are also discussed. As an example of the application of this methodology an analysis of the burnup reactivity credit for the three-dimensional model of the reactor RA spent fuel storage is described. (author)

  7. Tobacco use and nicotine dependence among the adult males of different socioeconomic groups within a medical college campus in Ammapettai, Kancheepuram district of South India

    Directory of Open Access Journals (Sweden)

    Glad M. I. Mohesh

    2014-06-01

    Full Text Available BACKGROUND: India is becoming one of the most tobacco consuming countries with the expected number of tobacco-related deaths equaling to more than 1.5 million annually by 2020. As strong socioeconomic differences exist in the pattern and type of tobacco consumption across large number of communities and through varied cultures across the country, the objective of this study is to understand if there is any difference by socioeconomic status (SES in nicotine dependence level and other characteristics related to tobacco quitting in tobacco users employed in a medical college campus. METHODS: Cross-sectional survey was conducted with a pretested questionnaire among 150 current tobacco users from different departments within the campus. The level of nicotine dependence was diagnosed using the Fagerstrom's nicotine dependence test. Kuppuswamys socioeconomic classification (2012 was used to group them by socioeconomic strata. Data was also collected on pattern and type of tobacco use, nicotine addiction, and attitude towards quitting tobacco use and tobacco control policies. Data collected was analysed using SPSS 21 and Epi-info 7 with the use of statistical tests of independence as well as binary logistic regression analysis. RESULTS: Nicotine dependence was higher in the lower SES groups than higher SES groups. Cigarette smoking prevailed among higher SES groups whereas bidi smoking was higher in lower SES groups. Smokeless tobacco was predominantly used in the lower SES groups. Awareness about the harmful effects of tobacco use was poorer in the low SES groups while the willingness to quit and the number of quit attempts did not differ by SES. More affluent tobacco users reported earlier initiation of tobacco use than those from lower SES groups. CONCLUSION: SES groups of tobacco users differ in type and pattern of tobacco use, dependence, age of initiation and awareness but not in intention and efforts to quit tobacco.

  8. 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%.

  9. Taking burnup credit for interim storage and transportation system for BWR fuels

    International Nuclear Information System (INIS)

    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)

  10. Separation of Molybdenum From Spent Fuel Solution in Burnup Measurements Process

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    In order to establish a kind of automatic radiochemistry separation procedure of nuclide 100Mo from spent fuel solution in burnup measurements process, a method of separating Mo quickly and effectively from the feed solution is needed. In the studies,

  11. 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.

  12. Economic viability to BeO-UO2 fuel burnup extension

    International Nuclear Information System (INIS)

    This paper presents the quantitative analysis results of research on the burnup effect on the nuclear fuel cycle cost of BeO-UO2 fuel. As a result of this analysis, if the burnup is 60 MWD/kg, which is the limit under South Korean regulations, the nuclear fuel cycle cost is 4.47 mills/kWh at 4.8wt% of Be content for the BeO-UO2 fuel. It is, however, reduced to 3.70 mills/kWh at 5.4wt% of Be content if the burnup is 75MWD/kg. Therefore, it seems very advantageous, in terms of the economic aspect, to develop BeO-UO2 fuel, which does not have any technical problem with its safety and is a high burnup and long life cycle nuclear fuel

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

    International Nuclear Information System (INIS)

    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

  14. U.S. Regulatory Research Program for Implementation of Burnup Credit in Transport Casks

    International Nuclear Information System (INIS)

    In 1999 the U.S. Nuclear Regulatory Commission (U.S. NRC) initiated a research program to support the development of technical bases and guidance that would facilitate the implementation of burnup credit into licensing activities for transport and dry cask storage. This paper reviews the following major areas of investigation: (1) specification of axial burnup profiles, (2) assumption on cooling time, (3) allowance for assemblies with fixed and removable neutron absorbers, (4) the need for a burnup margin for fuel with initial enrichments over 4 wt %, and (5) evaluation of assay data and critical experiments. The capabilities of a new computational tool that facilitates the performance and coupling of the depletion and criticality analyses needed for burnup credit are also discussed

  15. COREBN: A core burn-up calculation module for SRAC2006

    International Nuclear Information System (INIS)

    COREBN is an auxiliary code of the SRAC system for multi-dimensional core burn-up calculation based on the diffusion theory and interpolation of macroscopic cross-sections tabulated to local parameters such as burn-up degree, moderator temperature and so on. The macroscopic cross-sections are prepared by cell burn-up calculations with the collision probability method of SRAC. SRAC and COREBN have wide applicability for various types of cell and core geometries. They have been used mainly for the purpose of core burn-up management of research reactors in Japan Atomic Energy Agency. The report is a revision of the users manual for the latest version of COREBN served with the SRAC released in 2006. (author)

  16. Improving burnup performance of fast sodium cooled reactor by utilizing thorium based fuels

    International Nuclear Information System (INIS)

    To study the improvement of fuel burnup for fast reactors, thorium based fuels are investigated. In order to ensure the projected expansion of nuclear power is achieved in conjunction with reduced risk of nuclear weapons proliferation, new conventional sources of fuel will have to be made available. Thorium fuel cycles have many incentives such as the reduction of plutonium generation and consumption of LWR actinides, the provision of high performance burnup, and the conservation of 235U resources. This work examined the burnup reactivity loss and depletion analysis of thorium versus uranium based metal fuels. When compared the thorium based metallic fuel outperformed uranium based fuel with respect to higher actinide burnup and higher depletion rate of plutonium isotopes. (authors)

  17. Irradiation test for verification of PWR 48 GWd/t high burnup fuel

    International Nuclear Information System (INIS)

    Nuclear Power Engineering Corporation (NUPEC) has conducted the irradiation test for verification of the high burnup fuel performance under the sponsorship of the Ministry of Economy, Trade and Industry. (NUPEC-HB Project) As for PWR, the fuel burnup is extended by two steps. The Step I fuel (maximum fuel assembly discharge burnup: 48 GWd/t), has been utilized since 1989. And now, the preparation for the regular utilization of Step II fuel (maximum fuel assembly discharge burnup: 55 GWd/t), is being conducted. The results of pre- and post-irradiation tests on the Step I fuel irradiated in the Takahama-3 of Kansai Electric Power Co., Inc., were analyzed and evaluated. The irradiation performance of fuel rod, pellet, cladding and fuel assembly showed no remarkable difference compared with that of other published paper. Consequently the reliability and integrity of the Step I fuel was verified. (author)

  18. Implementation of burnup credit in spent fuel management systems. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    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

  19. Results of the isotopic concentrations of WWER calculation Burnup Credit Benchmark NO.2 (CB2)

    International Nuclear Information System (INIS)

    The purpose of this document is to present the results of the nuclide concentrations of the WWER Burnup Credit Benchmark NO.2 (CB2) that were performed in The Nuclear Technology Center of Cuba with available codes and libraries. The CB2 benchmark specification as the second phase of the WWER burnup credit benchmark is summarized in [1]. The CB2 benchmark focused on WWER burnup credit study proposed on the 97' Atomic Energy Research symposium [2]. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and cooling time. The depletion point 'ORIGEN2'[3] code was used for the calculation of the spent fuel concentration. This work also comprises the results obtained by other codes [4]. (Author)

  20. Fission gas release and fuel rod chemistry related to extended burnup

    International Nuclear Information System (INIS)

    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