MTR core loading pattern optimization using burnup dependent group constants
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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.
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
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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
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
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)
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.
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)
Considerations on burn-up dependent RIA and LOCA criteria
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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)
The dependence of the global neutronic parameters on the fuel burnup for CANDU SEU43 core
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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.)
Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1
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...
Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1
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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.
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)
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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
Development of burnup dependent fuel rod model in COBRA-TF
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
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)
Dependence of heavy metal burnup on nuclear data libraries for fast reactors
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...
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.
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
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
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
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.
ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for Candu Reactor Fuels
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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
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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)
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.)
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
Effects of axial burnup distributions on the reactivity of spent fuel
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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
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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)
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
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.
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
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
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
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)
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)
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)
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.
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.
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.
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)
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
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
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
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
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
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.)
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
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....
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
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)
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)
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
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)
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)
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
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)
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
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)
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)
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)
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.)
Detailed Burnup Calculations for Testing Nuclear Data
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
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
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
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
On the dimensional dependence of the electromagnetic duality groups
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.
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
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
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)
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
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)
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
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
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)
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)
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)
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).
Context dependent feature groups, a proposal for object representation.
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.
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.)
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.)
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
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)
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
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
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)
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.
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)
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
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)
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)
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.
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
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)
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
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
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.)
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.
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
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%.
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
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
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
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)
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
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
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)
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
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)
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
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)
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)
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)
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)
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)
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
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
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.
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
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 ...
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.
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
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)
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
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
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)
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
Effectiveness of Mindfulness-Based Group Therapy Compared to the Usual Opioid Dependence Treatment
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...
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.)
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
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
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)
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.)
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)
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)
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
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
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.
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.
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)
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
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)
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
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)
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)
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
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
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
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)
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).
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.)
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)
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)
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)
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)
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
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
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
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)
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.)
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
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)
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)
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...
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)
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
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 (k_{eff}) 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
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.
Assessing Colour-dependent Occupation Statistics Inferred from Galaxy Group Catalogues
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...
High-Resolution Crystal Structures Elucidate the Molecular Basis of Cholera Blood Group Dependence
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
Beyond the Sum of Parts: Voting with Groups of Dependent Entities.
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
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
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.
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
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
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)
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)
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)
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.
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)
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.
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.
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
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
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.
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
Group Interventions with Low-Income African American Women Recovering from Chemical Dependency.
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…
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…
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…
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
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
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)
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)
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)
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
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)
Kelvin-probe force microscopy of the pH-dependent charge of functional groups
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.
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
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
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.
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)
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
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
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.
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)
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)
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%.
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)
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
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
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
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)
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)
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)
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)
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
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.)
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)
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
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
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
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)
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)
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)
Thermonuclear burn-up in deuterated methane $CD_4$
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.
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)
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)
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
A questionnaire on survival of kittens depending on the blood groups of the parents.
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
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
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
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)
Salt effects on lamellar repeat distance depending on head groups of neutrally charged lipids.
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
A closely related group of RNA-dependent RNA polymerases from double-stranded RNA viruses.
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...
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.
The oxidative costs of reproduction are group-size dependent in a wild cooperative breeder.
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
Do the stellar populations of the brightest two group galaxies depend on the magnitude gap?
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, ...
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
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
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
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)
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)
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.
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)
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)
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)
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)
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)
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)
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)
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.
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)