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Sample records for burnup dependent group

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

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: Eleodor.nichita@uoit.ca [University of Ontario Institute of Technology, 2000 Simcoe Str. North, Oshawa, ON, Canada L1H 7K4 (Canada); Serghiuta, D.; Podobed, S. [Canadian Nuclear Safety Commission, 280 Slater Street, P.O. Box 1046 Station B, Ottawa, ON, Canada K1P 5S9 (Canada)

    2015-05-15

    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

  2. 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)

  3. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

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

  7. 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)

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

    CERN Document Server

    Ohki, S

    2003-01-01

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

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

    Science.gov (United States)

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

    2005-05-01

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

  10. 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.

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

    International Nuclear Information System (INIS)

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

  12. 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.)

  13. 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.

  14. 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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Ramamoorthy, Karthikeyan

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO2, UO2-Gd2O3, inhomogeneous MOX, and UO2-ThO2. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of 92U233-239, 93Np237-239, 94Pu238-243, 95Am241-244 (including isomers), and 96Cm242-245. Poisoning fission products are represented by 54Xe131,133,135, 48Cd113, 62Sm149,151,152, 64Gd154-160, 63Eu153,155, 36Kr83,85, 42Mo95, 43Tc99, 45Rh103, 47Ag109, 53I127,129,131, 55Cs133, 57La139, 59Pr141, 60Nd143-150, 61Pm147. Fission gases and volatiles included in the code are 36Kr83-86, 54Xe129-136, 52Te125-130, 53I127-131, 55Cs133-137, and 56Ba135-140. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-08-01

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

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

    Science.gov (United States)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

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

  3. 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.)

  4. 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

  5. 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)

  6. 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

  7. 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)

  8. Renormalization Scheme Dependence and Renormalization Group Summation

    CERN Document Server

    McKeon, D G C

    2016-01-01

    We consider logarithmic contributions to the free energy, instanton effective action and Laplace sum rules in QCD that are a consequence of radiative corrections. Upon summing these contributions by using the renormalization group, all dependence on the renormalization scale parameter mu cancels. The renormalization scheme dependence in these processes is examined, and a renormalization scheme is found in which the effect of higher order radiative corrections is absorbed by the behaviour of the running coupling.

  9. 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

  10. 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)

  11. 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.)

  12. 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)

  13. Detailed Burnup Calculations for Testing Nuclear Data

    Science.gov (United States)

    Leszczynski, F.

    2005-05-01

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

  14. 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

  15. High burnup in DIONISIO code

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-11-01

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

  17. 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)

  18. 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

  19. 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)

  20. 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

  1. 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.)

  2. The clustering of SDSS galaxy groups: mass and color dependence

    CERN Document Server

    Wang, Yu; Mo, H J; Bosch, Frank C van den; Weinmann, Simone W; Chu, Yaoquan

    2007-01-01

    We use a sample of galaxy groups selected from the SDSS DR 4 with an adaptive halo-based group finder to probe how the clustering strength of groups depends on their masses and colors. In particular, we determine the relative biases of groups of different masses, as well as that of groups with the same mass but with different colors. In agreement with previous studies, we find that more massive groups are more strongly clustered, and the inferred mass dependence of the halo bias is in good agreement with predictions for the $\\Lambda$CDM cosmology. Regarding the color dependence, we find that groups with red centrals are more strongly clustered than groups of the same mass but with blue centrals. Similar results are obtained when the color of a group is defined to be the total color of its member galaxies. The color dependence is more prominent in less massive groups and becomes insignificant in groups with masses $\\gta 10^{14}\\msunh$. We construct a mock galaxy redshift survey constructed from the large Mille...

  3. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-01

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

  5. 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%.

  6. 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.

  7. 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

  8. 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

  9. Calibration of burnup monitor installed in Rokkasho Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Naito, Hirofumi; Hirota, Masanari [Japan Nuclear Fuel Co. Ltd., Rokkasho, Aomori (Japan); Natsume, Koichiro [Isogo Engineering Center, Toshiba Corporation, Yokohama, Kanagawa (Japan); Kumanomido, Hironori [Nuclear Engineering Laboratory, Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2000-06-01

    Rokkasho Reprocessing Plant uses burnup credit for criticality control at the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. A burnup monitor measures nondestructively burnup value of a spent fuel assembly and guarantees the credit for burnup. For practical reasons, a standard radiation source is not used in calibration of the burnup monitor, but the burnup values of many spent fuel assemblies are measured based on operator-declared burnup values. This paper describes the concept of burnup credit, the burnup monitor, and the calibration method. It is concluded, from the results of calibration tests, that the calibration method is valid. (author)

  10. 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

  11. 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

  12. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. 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

  16. 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)

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

    OpenAIRE

    M. H. Altaf; Badrun, N. H.

    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 ...

  18. 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.

  19. 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

  20. 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.

  1. 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

  2. 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

  3. Burnup credit issues in transportation and storage

    International Nuclear Information System (INIS)

    Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the 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)

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-15

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

  6. 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)

  7. 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.

  8. 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.

  9. 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)

  10. 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)

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-02-01

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

  13. 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

  14. 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).

  15. 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.)

  16. 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

  17. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  20. 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.

  1. 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)

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

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

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

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

    Science.gov (United States)

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

    2016-10-01

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

  4. 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.

  5. 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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1996-05-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

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

  12. 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

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. 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.

  18. 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)

  19. 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

  20. 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.

  1. 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

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

    International Nuclear Information System (INIS)

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

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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)

  8. 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.)

  9. 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

  10. 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)

  11. 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)

  12. 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)

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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)

  18. 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)

  19. 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)

  20. 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)

  1. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    CERN Document Server

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

    2015-01-01

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

  5. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-15

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

  8. Calibration of burnup monitor in the Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oheda, K.; Naito, H.; Hirota, M. [Japan Nuclear Fuel Ltd., Aomori (Japan); Natsume, K. [Toshiba Corp., Yokohama, Kawasaki, Kanagawa (Japan); Kumanomido, H. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    1998-07-01

    The Rokkasho Reprocessing Plant has adopted a credit for burnup in criticality control in the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. The burnup monitor system, prepared for BWR and PWR type fuel assemblies, nondestructively measures the burnup value and determines the residual U-235 enrichment in a spent fuel assembly, and criticality is controlled by the value of residual U-235 enrichment in SFSF and by the value of top 50 cm average burnup in the Dissolution Facility. The burnup monitor consists of three measurement systems; a Boss gamma-ray profile measurement system, a high resolution gamma-ray spectrometry system, and a passive neutron measurement system. The monitor sensitivity is calibrated against operator-declared burnup values through repetitive measurements of 100 spent fuel assemblies: BWR 8 X 8, PWR 14 X 14. and 17 X 17. The outline of the measurement methods, objectives of the calibration, actual calibration method, and an example of calibration performed in a demonstration experiment are presented. (author)

  9. Research on irradiation behavior of superhigh burnup fuel

    International Nuclear Information System (INIS)

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

  10. Research on irradiation behavior of superhigh burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

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

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. 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.

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

    International Nuclear Information System (INIS)

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B4C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, keff, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The keff was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, keff was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2007-01-01

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

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

    Science.gov (United States)

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

    2003-01-01

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

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

    Science.gov (United States)

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

    2005-01-01

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

  1. Effect of burn-up and high burn-up structure on spent nuclear fuel alteration

    Energy Technology Data Exchange (ETDEWEB)

    Clarens, F.; Gonzalez-Robles, E.; Gimenez, F. J.; Casas, I.; Pablo, J. de; Serrano, D.; Wegen, D.; Glatz, J. P.; Martinez-Esparza, A.

    2009-07-01

    In this report the results of the experimental work carried out within the collaboration project between ITU-ENRESA-UPC/CTM on spent fuel (SF) covering the period 2005-2007 were presented. Studies on both RN release (Fast Release Fraction and matrix dissolution rate) and secondary phase formation were carried out by static and flow through experiments. Experiments were focussed on the study of the effect of BU with two PWR SF irradiated in commercial reactors with mean burn-ups of 48 and 60 MWd/KgU and; the effect of High Burn-up Structure (HBS) using powdered samples prepared from different radial positions. Additionally, two synthetic leaching solutions, bicarbonate and granitic bentonite ground wa ter were used. Higher releases were determined for RN from SF samples prepared from the center in comparison with the fuel from the periphery. However, within the studied range, no BU effect was observed. After one year of contact time, secondary phases were observed in batch experiments, covering the SF surface. Part of the work was performed for the Project NF-PRO of the European Commission 6th Framework Programme under contract no 2389. (Author)

  2. Multi-Group Formulation of the Temperature-Dependent Resonance Scattering Model and its Impact on Reactor Core Parameters

    Energy Technology Data Exchange (ETDEWEB)

    Ghrayeb, Shadi Z. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Mechanical and Nuclear Engineering; Ougouag, Abderrafi M. [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Ouisloumen, Mohamed [Westinghouse Electric Company, Cranberry Township, PA (United States); Ivanov, Kostadin N. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Mechanical and Nuclear Engineering

    2014-01-01

    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 accounts for it 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 the results done on this topic to date.

  3. Radionuclide Release from High Burnup Fuel

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. The commercial and technological impact of high burnup

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. Group classification of Schrödinger equations with position dependent mass

    Science.gov (United States)

    Nikitin, A. G.; Zasadko, T. M.

    2016-09-01

    Maximal kinematical invariance groups of the 2d Schrödinger equations with position dependent mass (PDM) and arbitrary potential are classified. It is demonstrated that there exist seven classes of such equations possessing non-equivalent continuous symmetry group. Three of these classes include arbitrary functions while the remaining ones are defined up to arbitrary parameters. In particular, for the case of a constant mass the class missing in the Boyer classification (Boyer 1974 Helv. Phys. Acta. 47 450) is indicated. A constructive test of (non)equivalence of a PDM system to a constant mass system is proposed.

  10. 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

  11. Mechanical Property Evaluation of High Burnup PHWR Fuel Clads

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

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

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

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  17. Aroma of Turmeric: Dependence on the Combination of Groups of Several Odor Constituents.

    Science.gov (United States)

    Hasegawa, Toshio; Nakatani, Kenta; Fujihara, Takashi; Yamada, Hideo

    2015-06-01

    Turmeric is a popular material that plays an important role in the flavor and fragrance industries. Although many compounds have been reported as components of turmeric, its aroma profile has not been clarified. Recently we have developed a new approach for evaluating the complex odors of materials based on recent research on the mechanism of odor recognition. Here we report the characteristic aroma properties of turmeric obtained through the investigation of its aroma profile. The hexane extract of turmeric had a turmeric-like odor, whereas the steam distillate of turmeric had a pungent, non-turmeric-like odor. We carried out bulb-to-bulb distillations of the extract and the steam distillate. For the hexane extract, two fractions with completely different odors were obtained. One was a high boiling point fraction (group A) with a turmeric-like odor, which consisted of ar-turmerone and β-turmerone as the main components, and the other was a low boiling point fraction (group B), which consisted of α-curcumene and β-sesquiphellandrene. In contrast, the bulb-to-bulb distillation of the steam distillate gave a fraction (group C) with a very different odor from groups A and B. Group C was composed of several kinds of alcohols that were not present in groups A and B. These results indicate that the group C fraction causes the different, pungent odor of the turmeric oil obtained by steam distillation. The variation in the aroma of turmeric depended on the combination of these three groups of odor constituents.

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

    Science.gov (United States)

    Axnér, Eva

    2014-10-01

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

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

    Science.gov (United States)

    Hishida, Mafumi; Yamamura, Yasuhisa; Saito, Kazuya

    2014-09-01

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

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

    International Nuclear Information System (INIS)

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

  1. Temperature dependence of C-terminal carboxylic group IR absorptions in the amide I' region.

    Science.gov (United States)

    Anderson, Benjamin A; Literati, Alex; Ball, Borden; Kubelka, Jan

    2015-01-01

    Studies of structural changes in peptides and proteins using IR spectroscopy often rely on subtle changes in the amide I' band as a function of temperature. However, these changes can be obscured by the overlap with other absorptions, namely the side-chain and terminal carboxylic groups. The former were the subject of our previous report (Anderson et al., 2014). In this paper we investigate the IR spectra of the asymmetric stretch of α-carboxylic groups for amino acids representing all major types (Gly, Ala, Val, Leu, Ser, Thr, Asp, Glu, Lys, Asn, His, Trp, Pro) as well as the C-terminal groups of three dipeptides (Gly-Gly, Gly-Ala, Ala-Gly) in D₂O at neutral pH. Experimental temperature dependent IR spectra were analyzed by fitting of both symmetric and asymmetric pseudo-Voigt functions. Qualitatively the spectra exhibit shifts to higher frequency, loss in intensity and narrowing with increased temperature, similar to that observed previously for the side-chain carboxylic groups of Asp. The observed dependence of the band parameters (frequency, intensity, width and shape) on temperature is in all cases linear: simple linear regression is therefore used to describe the spectral changes. The spectral parameters vary between individual amino acids and show systematic differences between the free amino acids and dipeptides, particularly in the absolute peak frequencies, but the temperature variations are comparable. The relative variations between the dipeptide spectral parameters are most sensitive to the C-terminal amino acid, and follow the trends observed in the free amino acid spectra. General rules for modeling the α-carboxylic IR absorption bands in peptides and proteins as the function of temperature are proposed.

  2. Temperature dependence of C-terminal carboxylic group IR absorptions in the amide I‧ region

    Science.gov (United States)

    Anderson, Benjamin A.; Literati, Alex; Ball, Borden; Kubelka, Jan

    2015-01-01

    Studies of structural changes in peptides and proteins using IR spectroscopy often rely on subtle changes in the amide I‧ band as a function of temperature. However, these changes can be obscured by the overlap with other absorptions, namely the side-chain and terminal carboxylic groups. The former were the subject of our previous report (Anderson et al., 2014). In this paper we investigate the IR spectra of the asymmetric stretch of α-carboxylic groups for amino acids representing all major types (Gly, Ala, Val, Leu, Ser, Thr, Asp, Glu, Lys, Asn, His, Trp, Pro) as well as the C-terminal groups of three dipeptides (Gly-Gly, Gly-Ala, Ala-Gly) in D2O at neutral pH. Experimental temperature dependent IR spectra were analyzed by fitting of both symmetric and asymmetric pseudo-Voigt functions. Qualitatively the spectra exhibit shifts to higher frequency, loss in intensity and narrowing with increased temperature, similar to that observed previously for the side-chain carboxylic groups of Asp. The observed dependence of the band parameters (frequency, intensity, width and shape) on temperature is in all cases linear: simple linear regression is therefore used to describe the spectral changes. The spectral parameters vary between individual amino acids and show systematic differences between the free amino acids and dipeptides, particularly in the absolute peak frequencies, but the temperature variations are comparable. The relative variations between the dipeptide spectral parameters are most sensitive to the C-terminal amino acid, and follow the trends observed in the free amino acid spectra. General rules for modeling the α-carboxylic IR absorption bands in peptides and proteins as the function of temperature are proposed.

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

    Science.gov (United States)

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

    2015-11-22

    Life-history theory assumes that reproduction entails a cost, and research on cooperatively breeding societies suggests that the cooperative sharing of workloads can reduce this cost. However, the physiological mechanisms that underpin both the costs of reproduction and the benefits of cooperation remain poorly understood. It has been hypothesized that reproductive costs may arise in part from oxidative stress, as reproductive investment may elevate exposure to reactive oxygen species, compromising survival and future reproduction and accelerating senescence. However, experimental evidence of oxidative costs of reproduction in the wild remains scarce. Here, we use a clutch-removal experiment to investigate the oxidative costs of reproduction in a wild cooperatively breeding bird, the white-browed sparrow weaver, Plocepasser mahali. Our results reveal costs of reproduction that are dependent on group size: relative to individuals in groups whose eggs were experimentally removed, individuals in groups that raised offspring experienced an associated cost (elevated oxidative damage and reduced body mass), but only if they were in small groups containing fewer or no helpers. Furthermore, during nestling provisioning, individuals that provisioned at higher rates showed greater within-individual declines in body mass and antioxidant protection. Our results provide rare experimental evidence that reproduction can negatively impact both oxidative status and body mass in the wild, and suggest that these costs can be mitigated in cooperative societies by the presence of additional helpers. These findings have implications for our understanding of the energetic and oxidative costs of reproduction, and the benefits of cooperation in animal societies. PMID:26582023

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

    CERN Document Server

    Trevisan, M; Khosroshahi, H G

    2016-01-01

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

  5. Need for higher fuel burnup at the Hatch Plant

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

    For several years, Monte Carlo burnup/depletion codes have appeared, which couple a Monte Carlo code to simulate the neutron transport to a deterministic method that computes the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3 dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the time-expensive Monte Carlo solver called at each time step. Therefore, great improvements in term of calculation time could be expected if one could get rid of Monte Carlo transport sequences. For example, it may seem interesting to run an initial Monte Carlo simulation only once, for the first time/burnup step, and then to use the concentration perturbation capability of the Monte Carlo code to replace the other time/burnup steps (the different burnup steps are seen like perturbations of the concentrations of the initial burnup step). This paper presents some advantages and limitations of this technique and preliminary results in terms of speed up and figure of merit. Finally, we will detail different possible calculation scheme based on that method. (authors)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. 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)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Alone against the group: A unanimously disagreeing group leads to conformity, but cardiovascular threat depends on one's goals.

    Science.gov (United States)

    Seery, Mark D; Gabriel, Shira; Lupien, Shannon P; Shimizu, Mitsuru

    2016-08-01

    A long history of research in psychology has studied the consequences of when individuals face a group that unanimously disagrees with them. However, relatively little research has attempted to understand individuals' internal reactions to such disagreement while it is experienced. Psychophysiological measures are particularly well suited for this purpose. We used the perspective of the biopsychosocial model of challenge/threat to test whether and under what circumstances expressing one's political opinion to a disagreeing group led to a cardiovascular threat response (high total peripheral resistance, low cardiac output). We hypothesized that, when participants were provided with a goal to fit in with the group, a disagreeing group would elicit cardiovascular responses consistent with greater threat than an agreeing group, but that this effect would disappear if not reverse when participants were provided with a goal to express their individuality. Results supported hypotheses and further revealed a divergence between cardiovascular responses and conformity behavior, such that a disagreeing group fostered conformity regardless of goal condition. These findings suggest that (a) facing the prospect of a disagreeing group need not necessarily result in the negative experience of threat (reflecting evaluating low resources/high demands), and (b) conformity behavior can mask a range of internal states.

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

    Science.gov (United States)

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

    2014-02-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Su, Bingjing; Hawari, Ayman, I.

    2004-03-30

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

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

    CERN Document Server

    Işık, Emre

    2015-01-01

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

  19. IL-33-Dependent Group 2 Innate Lymphoid Cells Promote Cutaneous Wound Healing.

    Science.gov (United States)

    Rak, Gregory D; Osborne, Lisa C; Siracusa, Mark C; Kim, Brian S; Wang, Kelvin; Bayat, Ardeshir; Artis, David; Volk, Susan W

    2016-02-01

    Breaches in the skin barrier initiate an inflammatory immune response that is critical for successful wound healing. Innate lymphoid cells (ILCs) are a recently identified population of immune cells that reside at epithelial barrier surfaces such as the skin, lung, and gut, and promote proinflammatory or epithelial repair functions after exposure to allergens, pathogens, or chemical irritants. However, the potential role of ILCs in regulating cutaneous wound healing remains undefined. Here, we demonstrate that cutaneous injury promotes an IL-33-dependent group 2 ILC (ILC2) response and that abrogation of this response impairs re-epithelialization and efficient wound closure. In addition, we provide evidence suggesting that an analogous ILC2 response is operational in acute wounds of human skin. Together, these results indicate that IL-33-responsive ILC2s are an important link between the cutaneous epithelium and the immune system, acting to promote the restoration of skin integrity after injury.

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

    International Nuclear Information System (INIS)

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The IAEA has recently completed two co-ordinated Research Programmes (CRPs) on The Development of Computer Models for Fuel Element Behaviour in Water Reactors, and on Fuel Modelling at Extended Burnup. Through these CRPs it became evident that there was a need to obtain data on fuel behaviour at high burnup. Data related o thermal behaviour, fission gas release and pellet to clad mechanical interaction were obtained and presented at the Technical Committee Meeting on Advances in Fuel Pellet Technology for Improved Performance at High Burnup which was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). The 34 papers from 10 countries are published in this proceedings and presented by a separate abstract. The papers were grouped in 6 sessions. First two sessions covered the fabrication of both UO2 fuel and additives and MOX fuel. Sessions 3 and 4 covered the thermal behaviour of both types of fuel. The remaining two sessions dealt with fission gas release and the mechanical aspects of pellet to clad interaction

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

    International Nuclear Information System (INIS)

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

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

    DEFF Research Database (Denmark)

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

    2009-01-01

    The conductance of π-conjugated molecular wires bonded to gold electrodes at zero bias is studied using density functional theory combined with nonequilibrium Green’s function method. For all systems considered, we find that the conductance length dependence follows the simple exponential law...... characteristic of tunneling through a barrier, G = Gc exp(−βL). For thiophene, pyrrole, and phenyl wires with thiol end-groups, we calculate decay constants (β) of 0.211, 0.257, and 0.264 Å−1, respectively, and contact conductances (Gc) of 1.25, 2.90, and 1.22G0, where G0 = 2e2/h is the conductance quantum....... In comparison, the corresponding values for amine-terminated thiophene are calculated to be β = 0.160 Å−1 and Gc = 0.038G0. These results show that (1) the contact resistance is mainly determined by the anchoring group and (2) the decay constant, which determines the conductance in the long wire limit...

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

    Directory of Open Access Journals (Sweden)

    Adem Pala

    2012-06-01

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

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

    Science.gov (United States)

    Ge, Tida; Zhu, Zhenke; Wu, Jinshui

    2016-04-01

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

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

    Science.gov (United States)

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

    2016-08-01

    Previous studies have found that when low-status group members are aware that their in-group is stereotyped as dependent by a specific out-group (i.e. a dependency meta-stereotype is salient), they are reluctant to seek help from the high-status out-group to avoid confirming the negative meta-stereotype. However, it is unclear whether low-status group members would seek more help in the context of a salient dependency meta-stereotype when there is low (vs. high) group boundary permeability. Therefore, we conducted two experiments to examine the moderating effect of permeability on meta-stereotype confirmation with a real group. In study 1, we manipulated the salience of the dependency meta-stereotype, measured participants' perceived permeability and examined their help-seeking behaviour in a real-world task. Participants who perceived low permeability sought more help when the meta-stereotype was salient (vs. not salient), whereas participants who perceived high permeability sought the same amount of help across conditions. In study 2, we manipulated the permeability levels and measured the dependency meta-stereotype. Participants who endorsed a high-dependency meta-stereotype sought more help than participants who endorsed a low-dependency meta-stereotype; this effect was particularly strong in the low-permeability condition. The implications of these results for social mobility and intergroup helping are discussed. PMID:25885332

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

    Institute of Scientific and Technical Information of China (English)

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

    2006-01-01

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

  9. Actinide-only burnup credit for spent fuel transport

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

  13. Comparison of neutron cross sections for selected fission products and isotopic composition analyses with burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Gil, C. S.; Kim, J. D.; Jang, J. H.; Lee, Y. D. [KAERI, Taejon (Korea)

    2003-10-01

    The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI-BNL international collaboration have been compared with the ENDF/B-VI release 7. Also, the influence of the new evaluations on isotopic compositions of the fission products as a function of burnup has been analyzed through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69 group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including new evaluations in resonance region covering thermal region, and ENDF/B-VII expected including those in upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows maximum difference of 4.78% compared to ENDF/B-VI.7. However, the isotopic compositions of all fission products calculated with ENDF/B-VII shows no differences compared to ENDF/B-VI.7.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

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

  16. New burnup calculation of TRIGA IPR-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  17. The REBUS experimental programme for burn-up credit

    International Nuclear Information System (INIS)

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

  18. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: nadiasam@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)

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

    International Nuclear Information System (INIS)

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

  20. Investigation of Burnup Credit Issues in BWR Fuel

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Adem Pala

    2012-06-01

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

  3. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

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

    Directory of Open Access Journals (Sweden)

    Michelle S. Ong

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Lee, Shinyoung; Kim, Heui-Baik

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Marco Pedrazzi

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

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

    International Nuclear Information System (INIS)

    The Technical Committee Meeting (TCM) included separate sessions on the specific topics of fuel thermal performance and fission product retention. On thermal performance, it is apparent that the capability exists to measure conductivity in high burnup fuel either by out-of-pile measurement or by instrumentation of test reactor rods. State-of-the-art modelling codes contain models for the conductivity degradation process, and hence adequate predictions of fuel temperature are achievable. Concerning fission product release, it is clear that many groups around the world are actively investigating the subject, with experimental and modelling programmes being pursued. However, a general consensus on the exact mechanisms of gas release and related gas bubble swelling has yet to emerge, even at medium burnup levels. Fission gas phenomena, not only the release to open volumes, but the whole sequence of processes taking place prior to this, need to be modelled in any modern fuel performance code. The presence of gaseous fission products may generate rapid fuel swelling during power transients, and this can cause PCI and rod failure. At high burnups, the quantity of released gases could give rise to pressures exceeding the safe limits. Modelling of pellet-cladding interaction (PCI) effects during transient operation is also an active area of study for many groups. In some situations a purely empirical approach to failure modelling can be justified, while for other applications a more detailed mechanistic approach is required. Another aspect of cladding modelling which was featured at the TCM concerned corrosion and hydriding. Although this issue can be the main life-limiting factor on fuel duty, it is apparent that modelling methods, and the experimental measurement techniques that underpin them, are adequate. A session was included on MOX fuel modelling. Substantial programmes of work, especially by the MOX vendors, appear to be underway to bring the level of understanding

  13. Burnup analysis of the power reactor, 1

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2008-01-01

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

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

    OpenAIRE

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-07-01

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

  20. Destructive radiochemical analysis of uraniumsilicide fuel for burnup determination

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

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

    NARCIS (Netherlands)

    van der Noort, V.

    2009-01-01

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

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

    Science.gov (United States)

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

    2012-01-01

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

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

    Science.gov (United States)

    Fischer, Ronald; Derham, Crysta

    2016-01-01

    We report a meta-analysis on the relationship between in-group bias and culture. Our focus is on whether broad macro-contextual variables influence the extent to which individuals favour their in-group. Data from 21,266 participants from 18 societies included in experimental and survey studies were available. Using Hofstede's (1980) and Schwartz (2006) culture-level predictors in a 3-level mixed-effects meta-analysis, we found strong support for the uncertainty-reduction hypothesis. An interaction between Autonomy and real vs artificial groups suggested that in low autonomy contexts, individuals show greater in-group bias for real groups. Implications for social identity theory and intergroup conflict are outlined.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. Evaluation of Isotopic Measurements and Burn-up Value of Sample GU3 of ARIANE Project

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Rodriguez Rivada, A.

    2014-07-01

    Estimation of the burn-up value of irradiated fuel and its isotopic composition are important for criticality analysis, spent fuel management and source term estimation. The practical way to estimate the irradiated fuel composition and burn.up value is calculation with validated code and nuclear data. Such validation of the neutronic codes and nuclear data requires the benchmarking with measured values. (Author)

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

    International Nuclear Information System (INIS)

    SCALE 4.3 is a modular code system designed for realizing standard computational analysis for licensing evaluation. Since now, spent fuel storage pools criticality analysis have been done considering this fuel as fresh, with its maximum enrichment. With burnup credit we can obtain cheaper and compact configurations. The procedure for calculating a spent fuel storage consists of a burnup calculation plus a criticality calculation. We can perform a conservative approximation for the burnup calculations using 1-D results, but, besides the geometry configurations for the 3-D criticality calculation. we need an appropriate approximation to model the burnup axial variation. We assume that for a burnup profile set, the most conservative profile is between the lower and the upper range of this profile, set. We consider only combinations of the maximum and minimum burnup in each axial region, for each burnup range. This gives an estimation of the different burnup shapes effect and the general characteristics of the most conservative shape. (Author) 6 refs

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

    International Nuclear Information System (INIS)

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

  11. Burnup Estimation for Plate Type Fuel Assembly Using SCALE6 Code

    Energy Technology Data Exchange (ETDEWEB)

    Alawneh, Luay M.; Park, Chang Je; Lee, Byungchul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    Accurate burnup estimation is not an easy job due to several reasons such as the effect of fission products and the power change caused by fuel refueling and depletion. The presence of fission products may distort the linear relationship between burnup and input parameters including power density and enrichment. The feasibility test of this approach has been done by comparing the results with a Monte Carlo code results. In this paper, it has been tried to get a crude formula to estimate burnup for an open pool type research reactor. In addition, we want to investigate the perturbation of each factor on burnup, and then combine the effects in one fitted formula for each cycle. This work is focused on calculating burnup for plate type fuel assembly of research reactors through a couple of code systems such as TRITON/NEWT and ORIGEN-ARP. Several sensitivity calculations have been done and the least square fitting is carried out to express a unified formula for burnup. The estimated burnup is compared with that of McCARD calculation. It is founded that the fitted burnup agrees well with the McCARD results.

  12. Development of advanced cladding material for burnup extension

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2015-01-01

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

  14. Calibration of burnup monitor of spent nuclear fuel installed at Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Matoba, Masaru; Wakabayashi, Genichiro [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Naito, Hirofumi; Hirota, Masanari [Nuclear Fuel Industries Ltd., Tokyo (Japan); Morizaki, Hidetoshi; Kumanomido, Hironori; Natsume, Koichiro [Toshiba Corp., Tokyo (Japan)

    2001-05-01

    The spent nuclear fuel storage pool of Rokkasho reprocessing plant adopts the burnup credit' conception. Spent fuel assemblies are measured every one by one, by burnup monitors, and stored to a storage rack which is designed with specified residual enrichment. For nuclear criticality control, it is necessary for the burnup monitor that the measured value includes a kind of margin, which consists of errors of the monitor. In this paper, we describe the error of the burnup monitors, and the way of taking of the margin. From the result of calibration of the burnup monitor carried out from July through November, 1999, we describe that the way of taking of the margin is validated. And comments about possibility of error reduction are remarked. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Preparation of data relevant to ''Equivalent Uniform Burnup'' and Equivalent Initial Enrichment'' for burnup credit evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan)

    2001-11-01

    Based on the PWR spent fuel composition data measured at JAERI, two kinds of simplified methods such as ''Equivalent Uniform Burnup'' and ''Equivalent Initial Enrichment'' have been introduced. And relevant evaluation curves have been prepared for criticality safety evaluation of spent fuel storage pool and transport casks, taking burnup of spent fuel into consideration. These simplified methods can be used to obtain an effective neutron multiplication factor for a spent fuel storage/transportation system by using the ORIGEN2.1 burnup code and the KENO-Va criticality code without considering axial burnup profile in spent fuel and other various factors introducing calculated errors. ''Equivalent Uniform Burnup'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis, in which the experimentally obtained isotopic composition together with a typical axial burnup profile and various factors such as irradiation history are considered on the conservative side. On the other hand, Equivalent Initial Enrichment'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis such as above when it is used in the so called fresh fuel assumption. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-01

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

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

    NARCIS (Netherlands)

    Martakis, Aias; Daneva, Maya; Wieringa, R.J.; Jean-Louis Cavarero, S.; Rolland, C.; Cavarero, J.-L.

    2013-01-01

    Agile practices on requirements dependencies are a relatively unexplored topic in literature. Empirical studies on it are scarce. This research sets out to uncover concepts that practitioners in companies of various sizes across the globe and in various industries, use for dealing with requirements

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

    International Nuclear Information System (INIS)

    UO2 pellets from ALUOX fuel elements were used in measuring the absorption coefficient of gamma radiation in UO2. The results of measurements of the energy dependence of the linear absorption coefficient (within 622 to 796 keV) and of the dependence on pellet density showed that in the given density interval the absorption coefficient was almost constant. The density interval was chosen to be typical for pellet fuel used in water cooled and water moderated power reactors. The results are also shown of the dependence of the mass absorption coefficient of gamma radiation in Zr on radiation energy and compared with the mass absorption coefficient of Mo; these also showed the independence of the absorption coefficient on density. The linear and mass absorption coefficients of UO2 are considerably high and correspond approximately to the absorption coefficient of lead. For the measured energy range the variation of absorption coefficient is about 40%, which causes errors in burnup determination. The efficiency was also determined of Ge(Li) detectors for the energy range 0.5 to 1.2 MeV. The determination of the above coefficients was used for improving the gamma fuel scanning technique in determining the activity and burnup of spent fuel elements. (J.P.)

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

    Science.gov (United States)

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

    2011-12-01

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

  4. Luminosity function of galaxies in groups in the SDSS DR7: the dependence on mass, environment and galaxy type

    CERN Document Server

    Zandivarez, Ariel

    2010-01-01

    We perform an analysis of the luminosities of galaxies in groups in the SDSS DR7. We analyse the luminosity function (LF) as a function of group mass for different photometric bands, galaxy populations, galaxy positions within the groups, and the group surrounding large scale density. We find that M* brightens and alpha becomes steeper as a function of mass in all SDSS photometric bands, in agreement with previous results. From the analysis of different galaxy populations, we observe that different methods to split galaxy populations, based on the concentration index or the colour-magnitude diagram, produce quite different behaviours in the luminosity trends, mainly for alpha. These discrepancies and the trends with mass mentioned above are explained when analysing the LF of galaxies classified simultaneously according to their concentrations and colours. We find that only the red spheroids have a LF that strongly depends on group mass. Late type galaxies, whether blue or red, have luminosity functions that d...

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

    Science.gov (United States)

    Washington, O

    1999-01-01

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

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

    Science.gov (United States)

    Washington, O

    1999-01-01

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

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

    Science.gov (United States)

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

    2007-01-01

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

  8. Anomalous length dependence of conductance of aromatic nanoribbons with amine anchoring groups

    KAUST Repository

    Bilić, Ante

    2012-09-06

    Two sets of aromatic nanoribbons, based around a common hexagonal scaffolding, with single and dual terminal amine groups have been considered as potential molecular wires in a junction formed by gold leads. Charge transport through the two-terminal device has been modeled using density functional theory (with and without self-interaction correction) and the nonequilibrium Green\\'s function method. The effects of wire length, multiple terminal contacts, and pathways across the junction have been investigated. For nanoribbons with the oligopyrene motif and conventional single amine terminal groups, an increase in the wire length causes an exponential drop in the conductance. In contrast, for the nanoribbons with the oligoperylene motif and dual amine anchoring groups the predicted conductance rises with the wire length over the whole range of investigated lengths. Only when the effects of self-interaction correction are taken into account, the conductance of the oligoperylene ribbons exhibits saturation for longer members of the series. The oligoperylene nanoribbons, with dual amine groups at both terminals, show the potential to fully harness the highly conjugated system of π molecular orbitals across the junction. © 2012 American Physical Society.

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

    Institute of Scientific and Technical Information of China (English)

    QU Chang-Zheng; ZHANG Shun-Li

    2004-01-01

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

  10. Group-size-dependent punishment of idle subordinates in a cooperative breeder where helpers pay to stay.

    Science.gov (United States)

    Fischer, Stefan; Zöttl, Markus; Groenewoud, Frank; Taborsky, Barbara

    2014-08-22

    In cooperative breeding systems, dominant breeders sometimes tolerate unrelated individuals even if they inflict costs on the dominants. According to the 'pay-to-stay' hypothesis, (i) subordinates can outweigh these costs by providing help and (ii) dominants should be able to enforce help by punishing subordinates that provide insufficient help. This requires that dominants can monitor helping and can recognize group members individually. In a field experiment, we tested whether cooperatively breeding cichlid Neolamprologus pulcher subordinates increase their help after a forced 'idle' period, how other group members respond to a previously idle helper, and how helper behaviour and group responses depend on group size. Previously, idle helpers increased their submissiveness and received more aggression than control helpers, suggesting that punishment occurred to enforce help. Subordinates in small groups increased their help more than those in large groups, despite receiving less aggression. When subordinates were temporarily removed, dominants in small groups were more likely to evict returning subordinates. Our results suggest that only in small groups do helpers face a latent threat of punishment by breeders as predicted by the pay-to-stay hypothesis. In large groups, cognitive constraints may prevent breeders from tracking the behaviour of a large number of helpers.

  11. Nursing dependency, diagnosis-related groups, and length of hospital stay

    OpenAIRE

    Halloran, Edward J.; Kiley, Marylou

    1987-01-01

    Most efforts to modify the diagnosis-related group (DRG) case classification system focus on variables related to medical management. In this study, we investigated the separate but related natures of medicine and nursing by examining 1,288 adult medical and surgical patients in an urban teaching hospital. The complexity of medical treatment was measured by use of the DRG relative cost weight. The nursing indicator was derived from a set of nursing diagnoses. We found that the DRG cost weight...

  12. Anomalous length dependence of conductance of aromatic nanoribbons with amine anchoring groups

    OpenAIRE

    Sanvito, Stefano

    2012-01-01

    PUBLISHED Two sets of aromatic nanoribbons, based around a common hexagonal scaffolding, with single and dual terminal amine groups have been considered as potential molecular wires in a junction formed by gold leads. Charge transport through the two-terminal device has been modeled using density functional theory (with and without self-interaction correction) and the nonequilibrium Green's function method. The effects of wire length, multiple terminal contacts, and pathways across the jun...

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

    International Nuclear Information System (INIS)

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

  14. Benefits of Group Foraging Depend on Prey Type in a Small Marine Predator, the Little Penguin

    Science.gov (United States)

    Sutton, Grace J.; Hoskins, Andrew J.; Arnould, John P. Y.

    2015-01-01

    Group foraging provides predators with advantages in over-powering prey larger than themselves or in aggregating small prey for efficient exploitation. For group-living predatory species, cooperative hunting strategies provide inclusive fitness benefits. However, for colonial-breeding predators, the benefit pay-offs of group foraging are less clear due to the potential for intra-specific competition. We used animal-borne cameras to determine the prey types, hunting strategies, and success of little penguins (Eudyptula minor), a small, colonial breeding air-breathing marine predator that has recently been shown to display extensive at-sea foraging associations with conspecifics. Regardless of prey type, little penguins had a higher probability of associating with conspecifics when hunting prey that were aggregated than when prey were solitary. In addition, success was greater when individuals hunted schooling rather than solitary prey. Surprisingly, however, success on schooling prey was similar or greater when individuals hunted on their own than when with conspecifics. These findings suggest individuals may be trading-off the energetic gains of solitary hunting for an increased probability of detecting prey within a spatially and temporally variable prey field by associating with conspecifics. PMID:26674073

  15. Benefits of Group Foraging Depend on Prey Type in a Small Marine Predator, the Little Penguin.

    Directory of Open Access Journals (Sweden)

    Grace J Sutton

    Full Text Available Group foraging provides predators with advantages in over-powering prey larger than themselves or in aggregating small prey for efficient exploitation. For group-living predatory species, cooperative hunting strategies provide inclusive fitness benefits. However, for colonial-breeding predators, the benefit pay-offs of group foraging are less clear due to the potential for intra-specific competition. We used animal-borne cameras to determine the prey types, hunting strategies, and success of little penguins (Eudyptula minor, a small, colonial breeding air-breathing marine predator that has recently been shown to display extensive at-sea foraging associations with conspecifics. Regardless of prey type, little penguins had a higher probability of associating with conspecifics when hunting prey that were aggregated than when prey were solitary. In addition, success was greater when individuals hunted schooling rather than solitary prey. Surprisingly, however, success on schooling prey was similar or greater when individuals hunted on their own than when with conspecifics. These findings suggest individuals may be trading-off the energetic gains of solitary hunting for an increased probability of detecting prey within a spatially and temporally variable prey field by associating with conspecifics.

  16. Benefits of Group Foraging Depend on Prey Type in a Small Marine Predator, the Little Penguin.

    Science.gov (United States)

    Sutton, Grace J; Hoskins, Andrew J; Arnould, John P Y

    2015-01-01

    Group foraging provides predators with advantages in over-powering prey larger than themselves or in aggregating small prey for efficient exploitation. For group-living predatory species, cooperative hunting strategies provide inclusive fitness benefits. However, for colonial-breeding predators, the benefit pay-offs of group foraging are less clear due to the potential for intra-specific competition. We used animal-borne cameras to determine the prey types, hunting strategies, and success of little penguins (Eudyptula minor), a small, colonial breeding air-breathing marine predator that has recently been shown to display extensive at-sea foraging associations with conspecifics. Regardless of prey type, little penguins had a higher probability of associating with conspecifics when hunting prey that were aggregated than when prey were solitary. In addition, success was greater when individuals hunted schooling rather than solitary prey. Surprisingly, however, success on schooling prey was similar or greater when individuals hunted on their own than when with conspecifics. These findings suggest individuals may be trading-off the energetic gains of solitary hunting for an increased probability of detecting prey within a spatially and temporally variable prey field by associating with conspecifics. PMID:26674073

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

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

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

  18. Simulation of triton burn-up in JET plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Loughlin, M.J.; Balet, B.; Jarvis, O.N.; Stubberfield, P.M. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    This paper presents the first triton burn-up calculations for JET plasmas using the transport code TRANSP. Four hot ion H-mode deuterium plasmas are studied. For these discharges, the 2.5 MeV emission rises rapidly and then collapses abruptly. This phenomenon is not fully understood but in each case the collapse phase is associated with a large impurity influx known as the ``carbon bloom``. The peak 14 MeV emission occurs at this time, somewhat later than that of the 2.5 MeV neutron peak. The present results give a clear indication that there are no significant departures from classical slowing down and spatial diffusion for tritons in JET plasmas. (authors). 7 refs., 3 figs., 1 tab.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-02-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Consideration of the depletion phenomena and isotopic uncertainties in burnup-credit criticality analysis places an increasing reliance on computational tools and significantly increases the overall complexity of the calculations. An automated analysis and data management capability is essential for practical implementation of large-scale burnup credit analyses that can be performed in a reasonable amount of time. STARBUCS is a new prototypic analysis sequence being developed for the SCALE code system to perform automated criticality calculations of spent fuel systems employing burnup credit. STARBUCS is designed to help analyze the dominant burnup credit phenomena including spatial burnup gradients and isotopic uncertainties. A search capability also allows STARBUCS to iterate to determine the spent fuel parameters (e.g., enrichment and burnup combinations) that result in a desired keff for a storage configuration. Although STARBUCS was developed to address the analysis needs for spent fuel transport and storage systems, it provides sufficient flexibility to allow virtually any configuration of spent fuel to be analyzed, such as storage pools and reprocessing operations. STARBUCS has been used extensively at Oak Ridge National Laboratory (ORNL) to study burnup credit phenomena in support of the NRC Research program

  2. Assessment of reactivity transient experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

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

    OpenAIRE

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

    2011-01-01

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

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

    Science.gov (United States)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bell, J.E.; Leone, A.; Bell, E.T.

    1986-05-01

    A glucosyltransferase, forming a predominantly al-6 linked glucan, was partially purified from the culture filtrate of S. mutans GS-5. The kinetic properties of the enzyme, assessed using the transfer of /sup 14/C glucose from sucrose into total glucan, were studied at pH values from pH 3.5 to 6.5. From the dependence of km on pH, a group with pKa = 5.5 must be protonated to maximize substrate binding. From plots of V/sub max/ vs pH two groups, with pKa's of 4.5 and 5.5 were indicated. The results suggest the involvement of either two carboxyl groups (one protonated, one unprotonated in the native enzyme) or a carboxyl group (unprotonated) and some other protonated group such as histidine, cysteine. Chemical modification studies showed that Diethylyrocarbonate (histidine specific) had no effect on enzyme activity while modification with p-phydroxy-mercuribenzoate or iodoacetic acid (sulfhydryl reactive) and carbodimide reagents (carboxyl specific) resulted in almost complete inactivation. Activity loss was dependent upon time of incubation and reagent concentration. The disaccharide lylose, (shown to be an inhibitor of the enzyme with similar affinity to sucrose) offers no protection against modification by the sulfhydryl reactive reagents.

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

    Energy Technology Data Exchange (ETDEWEB)

    Holly R. Trellue

    1998-12-01

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

  7. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

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

    Science.gov (United States)

    Isik, Emre

    2016-07-01

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

  9. Phase behavior of semifluorinated catanionic mixtures: head group dependence and spontaneous formation of vesicles.

    Science.gov (United States)

    Blanco, Elena; Olsson, Ulf; Ruso, Juan M; Schulz, Pablo C; Prieto, Gerardo; Sarmiento, Félix

    2009-03-15

    Hexadecyltrimethylammonium bromide (C(16)TAB)-sodium perfluorooctanoate (C(8)FONa) and hexadecylpyridynium bromide (C(16)PyB)-C(8)FONa catanionic semifluorinated mixtures have been studied by conductivity, dynamic light scattering (DLS), cryo-transmission electron microscopy (cryo-TEM) and polarizing microscopy. The regular solution theory, applicable for a limited fluorinated molar ratio, does not predict long-range electrostatic interactions. The results are consistent with the fact that in the hydrogenated-rich region the interaction is attractive in both catanionic mixtures. The systems containing pyridinium headgroups were of the stronger interaction. A transition from micelles was found in both mixtures as a function of fluorinated molar ratio. Special attention was devoted to the effect of the head group in the system properties. The information related with the mean vesicle radius measured by DLS was compared with the vesicle size distribution as well as the elastic properties of the bilayer measured with cryo-TEM.

  10. Evolutionary trails of plant group II Pyridoxal phosphate-dependent decarboxylase genes

    Directory of Open Access Journals (Sweden)

    Rahul Kumar

    2016-08-01

    Full Text Available Type II pyridoxal phosphate-dependent decarboxylase (PLP_deC enzymes play important metabolic roles during nitrogen metabolism. Recent evolutionary profiling of these genes revealed a sharp expansion of histidine decarboxylase (HDC genes in the members of Solanaceae family. In spite of the high sequence homology shared by PLP_deC orthologs, these enzymes display remarkable differences in their substrate specificities. Currently, limited information is available on the gene repertoires and substrate specificities of PLP_deCs which renders their precise annotation challenging and offers technical challenges in the immediate identification and biochemical characterization of their full gene complements in plants. Herein, we explored their evolutionary trails in a comprehensive manner by taking advantage of high-throughput data accessibility and computational approaches. We discussed the premise that has enabled an improved reconstruction of their evolutionary lineage and evaluated the factors offering constraints in their rapid functional characterization, till date. We envisage that the synthesized information herein would act as a catalyst for the rapid exploration of their biochemical specificity and physiological roles in more plant species.

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

    Science.gov (United States)

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

    2016-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

    We have looked at the interactions between polyamidoamine (PAMAM) dendrimers with different terminal groups (−COOH, −NH{sub 2}, −OH) and bovine insulin. The influence of PAMAM dendrimers on insulin was tested by measuring zeta potential and fluorescence quenching. The secondary structure of insulin in the presence of dendrimers was examined by circular dichroism. The effect of dendrimers on dithiotreitol-induced aggregation of insulin was investigated by spectrophotometry. Dendrimers quenched the fluorescence of insulin, but did not change its secondary structure. Thus dendrimers neither induce hormone aggregation nor inhibit the aggregation process induced by dithiotreitol (DTT), except at 0.01 µmol/l. Dendrimers–insulin interactions are mainly electrostatic. - Highlight: • The interactions between PAMAM dendrimers and insulin were investigated. • The PAMAM dendrimers can quench the fluorescence of insulin. • The PAMAM dendrimers did not change the secondary structure of insulin. • Dendrimers did not induce aggregation of hormone. • Dendrimers–insulin interaction is mainly electrostatic.

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

    International Nuclear Information System (INIS)

    We have looked at the interactions between polyamidoamine (PAMAM) dendrimers with different terminal groups (−COOH, −NH2, −OH) and bovine insulin. The influence of PAMAM dendrimers on insulin was tested by measuring zeta potential and fluorescence quenching. The secondary structure of insulin in the presence of dendrimers was examined by circular dichroism. The effect of dendrimers on dithiotreitol-induced aggregation of insulin was investigated by spectrophotometry. Dendrimers quenched the fluorescence of insulin, but did not change its secondary structure. Thus dendrimers neither induce hormone aggregation nor inhibit the aggregation process induced by dithiotreitol (DTT), except at 0.01 µmol/l. Dendrimers–insulin interactions are mainly electrostatic. - Highlight: • The interactions between PAMAM dendrimers and insulin were investigated. • The PAMAM dendrimers can quench the fluorescence of insulin. • The PAMAM dendrimers did not change the secondary structure of insulin. • Dendrimers did not induce aggregation of hormone. • Dendrimers–insulin interaction is mainly electrostatic

  14. Dependence of micelle size and shape on detergent alkyl chain length and head group.

    Directory of Open Access Journals (Sweden)

    Ryan C Oliver

    Full Text Available Micelle-forming detergents provide an amphipathic environment that can mimic lipid bilayers and are important tools for solubilizing membrane proteins for functional and structural investigations in vitro. However, the formation of a soluble protein-detergent complex (PDC currently relies on empirical screening of detergents, and a stable and functional PDC is often not obtained. To provide a foundation for systematic comparisons between the properties of the detergent micelle and the resulting PDC, a comprehensive set of detergents commonly used for membrane protein studies are systematically investigated. Using small-angle X-ray scattering (SAXS, micelle shapes and sizes are determined for phosphocholines with 10, 12, and 14 alkyl carbons, glucosides with 8, 9, and 10 alkyl carbons, maltosides with 8, 10, and 12 alkyl carbons, and lysophosphatidyl glycerols with 14 and 16 alkyl carbons. The SAXS profiles are well described by two-component ellipsoid models, with an electron rich outer shell corresponding to the detergent head groups and a less electron dense hydrophobic core composed of the alkyl chains. The minor axis of the elliptical micelle core from these models is constrained by the length of the alkyl chain, and increases by 1.2-1.5 Å per carbon addition to the alkyl chain. The major elliptical axis also increases with chain length; however, the ellipticity remains approximately constant for each detergent series. In addition, the aggregation number of these detergents increases by ∼16 monomers per micelle for each alkyl carbon added. The data provide a comprehensive view of the determinants of micelle shape and size and provide a baseline for correlating micelle properties with protein-detergent interactions.

  15. MCNPX Monte Carlo burnup simulations of the isotope correlation experiments in the NPP Obrigheim

    Energy Technology Data Exchange (ETDEWEB)

    Cao Yan, E-mail: ycao@anl.go [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Gohar, Yousry [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Broeders, Cornelis H.M. [Forschungszentrum Karlsruhe, Institute for Neutron Physics and Reactor Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2010-10-15

    This paper describes the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX Monte Carlo computer code package. The Monte Carlo simulation results are compared with the ICE-Experimental measurements for burnup up to 30 GWD/t. The comparison shows the good capabilities of the MCNPX computer code package for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR thermal reactor. The Monte Carlo simulation results show also good agreements with the experimental data for calculating several long-lived and stable fission products. However, for the americium and curium actinides, it is difficult to judge the predication capabilities for these actinides due to the large uncertainties in the ICE-Experimental data. In the MCNPX numerical simulations, a pin cell model is utilized to simulate the fuel lattice of the nuclear power reactor. Temperature dependent libraries based on JEFF3.1 nuclear data files are utilized for the calculations. In addition, temperature dependent libraries based ENDF/B-VII nuclear data files are utilized and the obtained results are very close to the JEFF3.1 results, except for {approx}10% differences in the prediction of the minor actinide isotopes buildup.

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

    Science.gov (United States)

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

    2016-08-01

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

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

    International Nuclear Information System (INIS)

    In order to establish a realistic burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of several modules such as TGBLA, ORIGEN, CITATION, MCNP, and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. A compact measurement system for a fuel assembly has been being developed to meet requirements for the burnup determination, the neutron emission-rate evaluation, and the nuclear materials management. For a spent MOX fuel, a neutron emission rate measurement method has been being developed. The system consists of Cd-Te detectors and / or fission chambers. Some model calculations were carried out for the latest design BWR fuels. The effect of taking burnup credit for a transportation cask is shown. (authors)

  18. Fission-gas release at extended burnups: effect of two-dimensional heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Yu, S.D. [Ryerson Polytechnic Univ., Toronto, Ontario (Canada); Lau, J.H.K

    2000-09-01

    To better simulate the performance of high-burnup CANDU fuel, a two-dimensional model for heat transfer between the pellet and the sheath has been added to the computer code ELESTRES. The model covers four relative orientations of the pellet and the sheath and their impacts on heat transfer and fission-gas release. The predictions of the code were compared to a database of 27 experimental irradiations involving extended burnups and normal burnups. The calculated values of fission gas release matched the measurements to an average of 94%. Thus, the two-dimensional heat transfer model increases the versatility of the ELESTRES code to better simulate fuels at normal as well as at extended burnups. (author)

  19. ThO{sub 2}-UO{sub 2} annular pins for high burnup fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caner, Marc; Dugan, Edward T

    2000-06-01

    The main purpose of this work is to investigate the use of annular fuel pins (particularly pins containing thorium dioxide) for high burnup fuel. The following parameters were evaluated and compared between postulated mixed thorium-uranium dioxide standard and annular (9% void fraction) type fuel assemblies, as a function of burnup: the infinite multiplication factor, the uranium and plutonium isotopic compositions, the fuel temperature coefficient of reactivity and the conversion ratio. We used the SCALE-4.3 code system. The calculation method consisted in obtaining actinide and fission product number densities as functions of assembly burnup, by means of a 1-D transport calculation combined with a 0-D burnup calculation. These number densities were then used in a 3-D Monte Carlo code for obtaining k{sub {infinity}} from two-dimensional-symmetry 'snapshots'.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

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

  4. Burnup of fusion produced tritons and 3He ions in PLT and PDX

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The purpose of the meeting was to review the state of the art in fission gas release and fuel rod chemistry related to extended burnup. The meeting was held in a time when national and international programmes on water reactor fuel irradiated in experimental reactors were still ongoing or had reached their conclusion, and when lead test assemblies had reached high burnup in power reactors and been examined. At the same time, several out-of-pile experiments on high burnup fuel or with simulated fuel were being carried out. As a result, significant progress has been registered since the last meeting, particularly in the evaluation of fuel temperature, the degradation of the global thermal conductivity with burnup and in the understanding of the impact on fission gas release. Fifty five participants from 16 countries and one international organization attended the meeting. 28 papers were presented. A separate abstract was prepared for each of the papers. Refs, figs, tabs and photos

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

    International Nuclear Information System (INIS)

    The purpose of this Technical Committee Meeting was to explore the status of international activities related to the use of burnup credit for spent fuel applications. This was the second major meeting on the issues of burnup credit for spent fuel management systems held since the IAEA began to monitor the uses of burnup credit in spent fuel management systems in 1997. Burnup credit for wet and dry storage systems is needed in many Member States to allow for increased initial fuel enrichment, and to increase the storage capacity and thus to avoid the need for extensive modifications of the spent fuel management systems involved. This document contains 31 individual papers presented at the Meeting; each of the papers was indexed separately

  7. Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages

    International Nuclear Information System (INIS)

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  8. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  9. Neutron Transport and Nuclear Burnup Analysis for the Laser Inertial Confinement Fusion-Fission Energy (LIFE) Engine

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E

    2008-10-24

    Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.

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

    Energy Technology Data Exchange (ETDEWEB)

    GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)] [and others

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

  12. A simple formula for local burnup based on constant relative reaction rate per nuclei

    CERN Document Server

    Yuan, Cenxi

    2015-01-01

    A simple and analytical formula is suggested to solve the problems on the local burnup and the isotope distributions. Present method considers that the slowing down neutrons going into the fuel rod is similar to the light going into the medium. Based on the assumption, the formula are obtained to calculate the reaction rates of $^{235}$U, $^{238}$U, and $^{239}$Pu and straightforward the local burnup and the isotope distributions. From a starting burnup point, the parameters of the formula are fitted to the reaction rates given by a Monte Carlo (MC) calculation. Then the present formula independently gives almost the same results as the MC calculation from the starting burnup point to high burnup, but takes just a few minutes. The relative reaction rate per nuclei are found to be almost independent on the radius (except $(n,\\gamma)$ of $^{238}$U) and burnup, providing a solid background for present formula. A combination of present formula and MC calculation is expected to have a nice balance on the accuracy ...

  13. A Criticality Evaluation of the GBC-32 Dry Storage Cask in PWR Burnup Credit

    International Nuclear Information System (INIS)

    The current criticality safety evaluation assumes the only unirradiated fresh fuels with the maximum enrichment in a dry storage cask (DSC) for conservatism without consideration of the depletion of fissile nuclides and the generation of neutron-absorbing fission products. However, the large conservatism leads to the significant increase of the storage casks required. Thus, the application of burnup credit which takes credit for the reduction of reactivity resulted from fuel depletion can increase the capacity in storage casks. On the other hand, the burnup credit application introduces lots of complexity into a criticality safety analysis such as the accurate estimation of the isotopic inventories and the burnup of UNFs and the validation of the criticality calculation. The criticality evaluation with an effect of burnup credit was performed for the DSC of GBC-32 by using SCALE 6.1/STARBUCS. keff values were calculated as a function of burnup and cooling time for four initial enrichments of 2, 3, 4, and 5 wt. % 235U. The values were calculated for the burnup range of 0 to 60,000 MWD/MTU, in increments of 10,000 MWD/MTU, and for five cooling times of 0, 5, 10, 20, and 40 years

  14. Water reactor fuel element modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The Technical Committee Meeting on Fuel Element Modelling at High Burnup and its Experimental Support was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its subject had been touched on in many of the IAEA's activities; however for the first time modellers and experimentalists were brought together to have an exchange of views on the research under way and to identify areas where new knowledge is necessary to improve the safety, reliability and/or economics of nuclear fuel. The timely organization of this meeting in conjunction with the second meeting of the Co-ordinated Research Programme on Fuel Modelling at Extended Burnup, in short ''FUMEX'', allowed fruitful participation of representatives of developing countries which are only rarely exposed to such a scientific event. The thirty-nine papers presented covered the status of codes and experimental facilities and the main phenomena affecting the fuel during irradiation, namely: thermal fuel performance, clad corrosion and pellet-cladding interaction (PCI) and fission gas release (FGR). Refs, figs, tabs

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

    CERN Document Server

    Tominaga, Nozomu; Blinnikov, Sergei I

    2015-01-01

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

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

    Science.gov (United States)

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

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

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

    Science.gov (United States)

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

    2016-07-01

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

  18. Dissolution of low burnup Fast Flux Test reactor fuel

    International Nuclear Information System (INIS)

    The first Fast-Flux Test Facility reactor fuel [mixed (U,Pu)O2 composition] has been used in dissolution tests for fuel reprocessing. The fuel tested here had a peak burnup of 0.22 at. %, with peak centerline temperatures of 19970C. Linear dissolution rates of 0.99 to 1.57 mm/h were determined for dissolver solution and fresh acid, respectively. Insoluble residues from dissolution at 950C ranged from 0.18 to 0.28% of the original fuel. From 2 to 37 wt % of the residue was recoverable plutonium. Dissolution at 290C yielded residues of 0.56 to 0.64% of the original fuel. The major elements present in the HF leached residue included Ru, Mo, and Rh. The recovered cladding from the 950C dissolution contained the equivalent of 198 mg of 239Pu per 100 g of hulls, while the cladding from the 290c experiments contained only 0.21 mg of 239Pu per 100 g of hulls. 9 references, 5 figures

  19. Miniature neutron source reactor burnup calculations using IRBURN code system

    International Nuclear Information System (INIS)

    Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Rob L [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Enercon Services, Inc.

    2011-03-14

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

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

    International Nuclear Information System (INIS)

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

  3. Tritium release from EXOTIC-7 orthosilicate pebbles. Effect of burnup and contact with beryllium during irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-03-01

    EXOTIC-7 was the first in-pile test with {sup 6}Li-enriched (50%) lithium orthosilicate (Li{sub 4}SiO{sub 4}) pebbles and with DEMO representative Li-burnup. Post irradiation examinations of the Li{sub 4}SiO{sub 4} have been performed at the Forschungszentrum Karlsruhe (FZK), mainly to investigate the tritium release kinetics as well as the effect of Li-burnup and/or contact with beryllium during irradiation. The release rate of Li{sub 4}SiO{sub 4} from pure Li{sub 4}SiO{sub 4} bed of capsule 28.1-1 is characterized by a broad main peak at about 400degC and by a smaller peak at about 800degC, and that from the mixed beds of capsule 28.2 and 26.2-1 shows again these two peaks, but most of the tritium is now released from the 800degC peak. This shift of release from low to high temperature may be due to the higher Li-burnup and/or due to contact with Be during irradiation. Due to the very difficult interpretation of the in-situ tritium release data, residence times have been estimated on the basis of the out-of-pile tests. The residence time for Li{sub 4}SiO{sub 4} from caps. 28.1-1 irradiated at 10% Li-burnup agrees quite well with that of the same material irradiated at Li-burnup lower than 3% in the EXOTIC-6 experiment. In spite of the observed shift in the release peaks from low to high temperature, also the residence time for Li{sub 4}SiO{sub 4} from caps. 26.2-1 irradiated at 13% Li-burnup agrees quite well with the data from EXOTIC-6 experiment. On the other hand, the residence time for Li{sub 4}SiO{sub 4} from caps. 28.2 (Li-burnup 18%) is about a factor 1.7-3.8 higher than that for caps. 26.2-1. Based on these data on can conclude that up to 13% Li-burnup neither the contact with beryllium nor the Li-burnup have a detrimental effect on the tritium release of Li{sub 4}SiO{sub 4} pebbles, but at 18% Li-burnup the residence time is increased by about a factor three. (J.P.N.)

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

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

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

  5. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    International Nuclear Information System (INIS)

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. Fifty-seven UO2, UO2/Gd2O3, and UO2/PuO2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on keff (which can be a function of the trending parameters) such that the biased keff, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading

  6. Practices and developments in spent fuel burnup credit applications. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency convened a technical committee Meeting on Requirements, Practices and Developments in Burnup Credit (BUC) Applications in Madrid, Spain, from 22 to 26 April 2002. The purpose of this meeting was to explore the progress and status of international activities related to the BUC applications for spent nuclear fuel. This meeting was the third major meeting on the uses of BUC for spent fuel management systems held since the IAEA began to monitor the uses of BUC in spent fuel management systems in 1997. The first major meeting was an Advisory Group meeting (AGM), which was held in Vienna, in October 1997. The second major meeting was a technical committee meeting (TCM), which was held in Vienna, in July 2000. Several consultants meetings were held since 1997 to advise and assist the IAEA in planning and conducting its BUC activities. The proceedings of the 1997 AGM were published as IAEA-TECDOC-1013, and the proceedings of the 2000 TCM as IAEA-TECDOC-1241. BUC for wet and dry storage systems, spent fuel transport, reprocessing and final disposal is needed in many Member States to allow for increased enrichment, and to increase storage capacities, cask capacities and dissolver capacities avoiding the need for extensive modifications. The use of BUC is a necessity for spent fuel disposal

  7. The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety

    International Nuclear Information System (INIS)

    It has been shown that supercriticality might occur for some postulated accident conditions at the TRIGA spent-fuel pool. However, the effect of burnup was not accounted for in previous studies. In this work, the combined effect of fuel burnup, pitch among fuel elements, and number of uniformly mixed absorber rods for a square arrangement on the spent-fuel pool keff is investigated.The Monte Carlo computer code MCNP4B with the ENDF-B/VI library and detailed three dimensional geometry was used. The WIMS-D code was used to model the isotopic composition of the standard TRIGA and FLIP fuel for 5, 10, 20 and 30% burnup level and 2- and 4-yr cooling time.The results show that out of the three studied effects, pitch from contact (3.75 cm) up to rack design pitch (8 cm), number of absorbers from zero to eight, 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 ∼0.4 for standard and ∼0.3 for FLIP fuel. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g., for contact of standard fuel elements with eight absorber rods among them, keff values are smaller for ∼0.2 (∼0.1 for FLIP) than for arrangements without absorber rods almost regardless of the burnup. The effect of burnup is the smallest. For standard fuel elements, it is ∼0.1 for almost all pitches and numbers of absorbers. For FLIP fuel, it is smaller for a factor of 3, but increases with the burnup for compact arrangements. Cooling time of fuel has just a minor effect on the keff of spent-fuel pool and can be neglected in spent-fuel pool design

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

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R.O.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L. [Pacific Northwest Laboratory, Richland, WA (United States)

    1996-03-01

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

  10. A Simple Formula for Local Burnup and Isotope Distributions Based on Approximately Constant Relative Reaction Rate

    Directory of Open Access Journals (Sweden)

    Cenxi Yuan

    2016-01-01

    Full Text Available A simple and analytical formula is suggested to solve the problems of the local burnup and the isotope distributions. The present method considers two extreme conditions of neutrons penetrating the fuel rod. Based on these considerations, the formula is obtained to calculate the reaction rates of 235U, 238U, and 239Pu and straightforward the local burnup and the isotope distributions. Starting from an initial burnup level, the parameters of the formula are fitted to the reaction rates given by a Monte Carlo (MC calculation. Then the present formula independently gives very similar results to the MC calculation from the starting to high burnup level but takes just a few minutes. The relative reaction rates are found to be almost independent of the radius (except (n,γ of  238U and the burnup, providing a solid background for the present formula. A more realistic examination is also performed when the fuel rods locate in an assembly. A combination of the present formula and the MC calculation is expected to have a nice balance between the numerical accuracy and time consumption.

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

    Science.gov (United States)

    Kolstad, Erik; Vitanza, Carlo

    1992-06-01

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

  12. A simplified burnup calculation strategy with refueling in static molten salt reactor

    International Nuclear Information System (INIS)

    Molten Salt Reactors, by nature can be refuelled and reprocessed online. Thus, a simulation methodology has to be developed which can consider online refueling and reprocessing aspect of the reactor. To cater such needs a simplified burnup calculation strategy to account for refueling and removal of molten salt fuel at any desired burnup has been identified in static molten salt reactor in batch mode as a first step of way forward. The features of in-house code ITRAN has been explored for such calculations. The code also enables us to estimate the reactivity introduced in the system due to removal of any number of considered nuclides at any burnup. The effect of refueling fresh fuel and removal of burned fuel has been studied in batch mode with in-house code ITRAN. The effect of refueling and burnup on change in reactivity per day has been analyzed. The analysis of removal of 233Pa at a particular burnup has been carried out. The similar analysis has been performed for some other nuclides also. (author)

  13. Kinetic parameter calculation as function of burn-up of candu reactor

    International Nuclear Information System (INIS)

    Kinetic parameter calculation as function of burn-up of candu reactor. Kinetic marameter calculation as function of burp-up of CANDU reactor with Canflex fuel type-CANDU has been done. This type of fuel is currently being develop, so kinetic parameter such as effective delay neutron fraction (.......), delay neutron decay constant ( .... ) and prompt neutron generation time ( ...... ) are very important for analysis of reactor operation safety. WIMS-CRNL code was used to generate macroscopic cross section and reaction rate based on transport theory. Fast and thermal neutron velocity and macroscopic cross section fission product of the unit cell were determined by KINETIC Code. The result of calculation showed that the value of effective delay neutron fraction was 7,785616 x 10-3 at the beginning of operation at burn-up of 0 MWD/T and after the reactor operated at burn-up of 7,2231 x 10-3 MWD/T was 4,962766 x 10-3, or reduced by 36%. The value of prompt generation time was 9,982703 x 10-4 s at the beginning of operation at burn-up of 0 MWD/T and 8,965416 x 10-4 s after the reactor operated at burn-up of 7,2231 x 103 MWD/T, or reduced by 10%. The result of calculation showed that the values of effective delay neutron fraction and prompt neutron generation time are still great enough

  14. Results of the isotopic concentrations of VVER calculational burnup credit benchmark no. 2(cb2

    International Nuclear Information System (INIS)

    The characterization of the irradiated fuel materials is becoming more important with the Increasing use of nuclear energy in the world. The purpose of this document is to present the results of the nuclide concentrations calculated Using Calculation VVER Burnup Credit Benchmark No. 2(CB2). The calculations were Performed in The Nuclear Technology Center of Cuba. The CB2 benchmark specification as the second phase of the VVER burnup credit benchmark is Summarized in [1]. The CB2 benchmark focused on VVER burnup credit study proposed on the 97' AER Symposium [2]. It should provide a comparison of the ability of various code systems And data libraries to predict VVER-440 spent fuel isotopes (isotopic concentrations) using Depletion analysis. This phase of the benchmark calculations is still in progress. CB2 should be finished by summer 1999 and evaluated results could be presented on the next AER Symposium. The obtained results are isotopic concentrations of spent fuel as a function of the burnup and Cooling time. The depletion point ORIGEN2[3] code was used for the calculation of the spent Fuel concentration. The depletion analysis was performed using the VVER-440 irradiated fuel assemblies with in-core Irradiation time of 3 years, burnup of the 30000 mwd/TU, and an after discharge cooling Time of 0 and 1 year. This work also comprises the results obtained by other codes[4].

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

    Science.gov (United States)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

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

  16. Investigation of research and development subjects for very high burnup fuel. Development of fuel cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Nagase, Fumihisa; Suzuki, Masahide; Furuta, Teruo; Suzuki, Yasufumi; Hayashi, Kimio; Amano, Hidetoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1993-05-01

    Plutonium use as well as burnup extension of UO{sub 2} fuel is an important subject for the strategy of utilization of the nuclear energy in LWRs. A higher burnup is favorable to MOX fuel in economic respect and for effective use of plutonium. Therefore, the concept of a `very high burnup` aiming at the maximum bundle burnup of 100GWd/t has been proposed assuming use of MOX fuel. The authors have investigated research and development subjects for the fuel pellet and the cladding material to be developed. The present report shows the results on the cladding material. In order to achieve a very high burnup, development of the cladding material with higher corrosion and radiation resistance compared with Zircaloy is necessary. In this report, zirconium based alloy, stainless steel, nickel and titanium based alloys, ceramics, etc. were reviewed considering water corrosion resistance, thermal and mechanical properties, radiation effects, etc. Furthermore, capability of these materials as the fuel cladding was discussed focusing on water side corrosion and radiation effect on mechanical properties. As a result, candidate materials at present and the required research tasks were shown with issues for the development. (author) 66 refs.

  17. Investigation of research and development subjects for very high burnup fuel

    International Nuclear Information System (INIS)

    Plutonium use as well as burnup extension of UO2 fuel is an important subject for the strategy of utilization of the nuclear energy in LWRs. A higher burnup is favorable to MOX fuel in economic respect and for effective use of plutonium. Therefore, the concept of a 'very high burnup' aiming at the maximum bundle burnup of 100GWd/t has been proposed assuming use of MOX fuel. The authors have investigated research and development subjects for the fuel pellet and the cladding material to be developed. The present report shows the results on the cladding material. In order to achieve a very high burnup, development of the cladding material with higher corrosion and radiation resistance compared with Zircaloy is necessary. In this report, zirconium based alloy, stainless steel, nickel and titanium based alloys, ceramics, etc. were reviewed considering water corrosion resistance, thermal and mechanical properties, radiation effects, etc. Furthermore, capability of these materials as the fuel cladding was discussed focusing on water side corrosion and radiation effect on mechanical properties. As a result, candidate materials at present and the required research tasks were shown with issues for the development. (author) 66 refs

  18. Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model

    Directory of Open Access Journals (Sweden)

    Abdul Waris

    2008-03-01

    Full Text Available Plutonium (Pu and minor actinides (MA recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.

  19. TIME-DEPENDENT MULTI-GROUP MULTI-DIMENSIONAL RELATIVISTIC RADIATIVE TRANSFER CODE BASED ON SPHERICAL HARMONIC DISCRETE ORDINATE METHOD

    Energy Technology Data Exchange (ETDEWEB)

    Tominaga, Nozomu; Shibata, Sanshiro [Department of Physics, Faculty of Science and Engineering, Konan University, 8-9-1 Okamoto, Kobe, Hyogo 658-8501 (Japan); Blinnikov, Sergei I., E-mail: tominaga@konan-u.ac.jp, E-mail: sshibata@post.kek.jp, E-mail: Sergei.Blinnikov@itep.ru [Kavli Institute for the Physics and Mathematics of the Universe (WPI), The University of Tokyo, 5-1-5 Kashiwanoha, Kashiwa, Chiba 277-8583 (Japan)

    2015-08-15

    We develop a time-dependent, multi-group, multi-dimensional relativistic radiative transfer code, which is required to numerically investigate radiation from relativistic fluids that are involved in, e.g., gamma-ray bursts and active galactic nuclei. The code is based on the spherical harmonic discrete ordinate method (SHDOM) which evaluates a source function including anisotropic scattering in spherical harmonics and implicitly solves the static radiative transfer equation with ray tracing in discrete ordinates. We implement treatments of time dependence, multi-frequency bins, Lorentz transformation, and elastic Thomson and inelastic Compton scattering to the publicly available SHDOM code. Our code adopts a mixed-frame approach; the source function is evaluated in the comoving frame, whereas the radiative transfer equation is solved in the laboratory frame. This implementation is validated using various test problems and comparisons with the results from a relativistic Monte Carlo code. These validations confirm that the code correctly calculates the intensity and its evolution in the computational domain. The code enables us to obtain an Eddington tensor that relates the first and third moments of intensity (energy density and radiation pressure) and is frequently used as a closure relation in radiation hydrodynamics calculations.

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

    Science.gov (United States)

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

    2016-05-15

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

  1. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel[Dissertation 17527

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

    , relevant Xe isotope concentrations in fission gas inclusions have not been reported yet. The aim of this work was focused on quantitative analysis of Xe isotope concentrations in inclusions of high burn-up samples to determine gas pressures within fuel pores. Therefore, various calibration strategies were studied in great detail. The investigation of Xe-implantation showed that the uncertainties in Xe ion implantation are too high to produce suitable Xe-implanted standard materials. The generation of Xe-bubbles in Plexiglass was also investigated and Xe inclusions of different bubble sizes were generated. However, the size determinations of the inclusions, the opening and possible gas losses through the Plexiglass were factors preventing accurate Xe quantification. The use of direct Xe gas injection was considered as a suitable strategy for establishing calibration curves. The Xe response was studied in dependence on gas concentration, gas volume and presence or absence of matrix in the ICP (Inductively Coupled Plasma). To investigate the ionization behavior of Xe in the ICP, the ablation behavior of ZrO{sub 2} and UO{sub 2} was determined in dependence on different wavelengths (266 nm, 193 nm), crater diameters, repetition rates and carrier gases (Ar, He). The mass ablated per laser pulse was very similar for the two matrices. The simultaneous introduction of matrix and gas into the ICP showed insignificant matrix effects. Even the absence of matrix effects was shown, filtering of the matrix before entering the plasma was evaluated to be crucial. Applying a filter leads to a complete gas/solid separation and opens access to the matrix interfering {sup 134}Xe and {sup 136}Xe isotopes. Thus, calibration gas addition in combination with an aerosol filter was the key strategy for quantification of Xe fission gas production in high burn-up fuel samples. Gas addition in volumes of 1 to 100 {mu}l and at a concentration of 20 and 200 {mu}g/g provided linear calibration curves

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

    Science.gov (United States)

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

    2013-04-01

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

  3. Irradiation behavior of FBTR mixed carbide fuel at various burn-ups

    International Nuclear Information System (INIS)

    The fast breeder test reactor at Kalpakkam has completed nearly 25 years of operation and is now operating at 18 MWt capacity with 46 fuel subassemblies (FSA) in the core consisting of 27 Mark-I (70% PuC + 30% UC), 13 Mark-II (55% PuC + 45% UC) and 6 MOX (44% PuO2 + 56% UO2) and one test PFBR FSA. Post Irradiation Examination (PIE) campaigns on FSAs at different burnup levels has provided valuable information about the irradiation behavior of the carbide fuel. This paper gives a summary of the irradiation performance of the carbide fuel evaluated through some of the investigations such as neutron radiography, x-radiography, gamma scanning, fission gas analysis and ceramography. Burnup of the carbide fuel could be enhanced from the initial design burnup limit of 50 GWd/t to 165 GWd/through systematic PIE. (author)

  4. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Wang Tienko E-mail: tkw@faculty.nthu.edu.tw; Peir Jinnjer

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products {sup 97}Zr/{sup 97}Nb, {sup 132}I, and {sup 140}La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by re irradiating the spent fuels in a nuclear reactor. Based on the measured activities, {sup 235}U burn-up values can be deduced by iterative calculations. The complication caused by {sup 239}Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products {sup 137}Cs, {sup 134}Cs/{sup 137}Cs ratio and {sup 106}Ru/{sup 137}Cs ratio.

  5. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry

    International Nuclear Information System (INIS)

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by re irradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio

  6. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry.

    Science.gov (United States)

    Wang, T K; Peir, J J

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by reirradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio. PMID:10670930

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

    Science.gov (United States)

    Sloma, Tanya Noel

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

  8. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  9. Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor

    International Nuclear Information System (INIS)

    The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)

  10. Criticality evaluation of high density spent fuel storge rack under normal condition using burnup credit

    International Nuclear Information System (INIS)

    The high density spent fuel storage rack Boraflex was known to experience changes of its physical property and to dissolve under exposure to radiation in an aqueous environment for long period of time. In this study, the criticality evaluation for spent fuel storage rack of Ulchin Unit 2 under normal condition was performed assuming complete loss of 10B from the Boraflex and applying burnup credit. Criticality evaluation code KENO-V.a. from SCALE4.4 system was benchmarked against critical experiments to obtain the calculation bias and bias uncertainties. The manufacturing tolerances of nuclear fuel and storage rack and their reactivity uncertainties were derived, as well. Considering those bias and uncertainties of calculation, the criticality of spent fuel storage under normal condition was conservatively evaluated. The criticality evaluation result using burnup credit can be presented as a spent fuel loading curve that indicates the acceptable burnup domain in spent fuel storage pool. The spent fuels with various initial enrichments and discharge fuel burnup can be safely accommodated in the storage without taking any boron credit from Boraflex, provided the combination falls within the acceptable domain in the loading curve. The spent fuel with initial enrichment of 5.0w/o was evaluated to meet the subcritical safety if its burnup is over 43.0GWD/MTU. The criticality evaluation result also showed that spent fuels with the initial enrichment less than 1.6w/o were able to be stored in the storage pool regardless of their burnup. Conclusively, in the Region 2 of the spent fuel storage pool, the maximum keff , considering all uncertainties, was calculated as 0.94818

  11. Estimating Burnup for UMo Plate Type Fuel with Least Square Fitting

    Energy Technology Data Exchange (ETDEWEB)

    Alawneh, Luay M.; Jaradat, Mustafa K. [Univ. of Science and Technology, Daejeon (Korea, Republic of); Park, Chang Je; Lee, Byungchul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The feasibility test of this approach has been done by comparing the results with a Monte Carlo code results. UMo fuel is a promising candidate for a high performance research reactor and provides better fuel performance including an extended burnup and swelling resistance. Additionally, its relatively high uranium content provides high power density. However, when irradiating UMo fuel in the core, lots of pores are produced due to an extensive interaction between the UMo and Al matrix. The pore leads to an expansion of fuel meat and may result in a fuel failure after all. This problem has almost been solved by using an optimal Si additive to depress the interaction layer. An international program has been performed to manufacture a robust UMo fuel. However, in terms of neutronics, the absorption cross section of Mo is much higher than that of Si, and thus a slightly high uranium density of UMo fuel is required to provide equivalent characteristics to U{sub 3}Si{sub 2} fuel. Recently, Korea considers U-Mo fuel for the KJRR design, which is under design stage. This work is focused on calculating burnup for plate type UMo fuel through a couple of code systems such as TRITON/NEWT and ORIGEN-ARP. The estimated burnup is compared with that of MCNPX calculation. It is founded that the fitted burnup agrees well with the MCNPX results. This approach will be applicable to easily estimate discharge burnup in research reactor without additional burden. However, some sensitivity tests required for another parameters in order to obtain burnup exactly.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. In-core fuel management amd attainable fuel burn-up in TRIGA

    International Nuclear Information System (INIS)

    The principles of in-core fuel management in research reactors, and especially in TRIGA, are discussed. Calculations made to determine the attainable fuel burn-up values of various fuel element types in the Otaniemi TRIGA Mark II reactor are described and the results obtained are given. Recommendations are given of how to perform the in-core fuel management to achieve good fuel utilization. The results obtained indicate that burn-up values of up to 5 and 2.5 MWd/element can be achieved for the 8 wt-% U Al clad and the 8.5 wt-% U SS clad elements, respectively. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Ishijima, Kiyomi; Fuketa, Toyoshi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-01-01

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

  16. The use of burnup credit in criticality control for the Korean spent fuel management program

    International Nuclear Information System (INIS)

    More than 25% k-eff saving effect is observed in this burnup credit analysis. This mainly comes from the adoption of actinide nuclides and fission products in the criticality analysis. By taking burnup credit, the high capacity of the storage and transportation can be more fully utilized, reducing the space of storage and the number of shipments. Larger storage and fewer shipments for a given inventory of spent fuel result should in remarkable cost savings and more importantly reduce the risks to the public and occupational workers for the Korean Spent Fuel Management Program

  17. Post Irradiation Examination Plan for High-Burnup Demonstration Project Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    This test plan describes the experimental work to be implemented by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) to characterize high burnup (HBU) spent nuclear fuel (SNF) in conjunction with the High Burnup Dry Storage Cask Research and Development Project and serves to coordinate and integrate the multi-year experimental program to collect and develop data regarding the continued storage and eventual transport of HBU (i.e., >45 GWd/MTU) SNF. The work scope involves the development, performance, technical integration, and oversight of measurements and collection of relevant data, guided by analyses and demonstration of need.

  18. Extended burnup demonstration: reactor fuel program. Pre-irradiation characterization and summary of pre-program poolside examinations. Big Rock Point extended burnup fuel

    International Nuclear Information System (INIS)

    This report is a resource document characterizing the 64 fuel rods being irradiated at the Big Rock Point reactor as part of the Extended Burnup Demonstration being sponsored jointly by the US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities. The program entails extending the exposure of standard BWR fuel to a discharge average of 38,000 MWD/MTU to demonstrate the feasibility of operating fuel of standard design to levels significantly above current limits. The fabrication characteristics of the Big Rock Point EBD fuel are presented along with measurement of rod length, rod diameter, pellet stack height, and fuel rod withdrawal force taken at poolside at burnups up to 26,200 MWD/MTU. A review of the fuel examination data indicates no performance characteristics which might restrict the continued irradiation of the fuel

  19. Effects of fuel burn-up and cooling periods on thermal responses in a repository for spent nuclear fuels

    International Nuclear Information System (INIS)

    Flexible methods and codes have been developed to calculate thermal responses in a spent nuclear fuel repository located in granitic bedrock. The thermal resistance of the backfill material makes an essential contribution to the temperature at the canister surface. The backfill is however so thin compared to the rock masses of the repository that a stationary solution for the temperature drop across the backfill can be used. Combining analytical solutions of thermal diffusion in the rock and the stationary temperature following the released heat power across the backfill gives a very good description of the thermal behaviour of a repository. Temperatures at different points in a repository were calculated for encapsuled spent fuel with different cooling times (10, 20, 30 and 40 years) and having different burn-up values (33, 35 and 45 MWd/kgU). The amount of spent fuel in each canister was supposed to originate from 1.4 tons of fresh Uranium in the calculated examples. A more crucial variable in a thermal analysis of a repository configuration is however the heat power of canisters as a function of time and not the content of canisters. The study shows that the largest temperature rise at the boundary of a centrally located canister and buffer mass will be at most 85 K with assumed thermal data, when the heat power of 900 canisters does not exceed 1200-1400 W each at the time of emplacement depending on burn-up and cooling time. Assuming a higher value for the heat conductivity of the buffer material corresponding to that of the water saturated bentonite gives 15% higher limits respectively. The canisters were supposed to be placed in 15 rows 25 m apart from each other and 60 canisters in each row. The distance between canisters was taken to be 6 m

  20. First-Principles Momentum Dependent Local Ansatz Approach to the Ground-State Properties of Iron-Group Transition Metals

    Science.gov (United States)

    Kakehashi, Yoshiro; Chandra, Sumal

    2016-08-01

    The ground-state properties of iron-group transition metals from Sc to Cu have been investigated on the basis of the first-principles momentum dependent local ansatz (MLA) theory. Correlation energy gain is found to show large values for Mn and Fe: 0.090 Ry (Mn) and 0.094 Ry (Fe). The Hund-rule coupling energies are found to be 3000 K (Fe), 1400 K (Co), and 300 K (Ni). It is suggested that these values can resolve the inconsistency in magnetic energy between the density functional theory and the first-principles dynamical coherent potential approximation theory at finite temperatures. Charge fluctuations are shown to be suppressed by the intra-orbital correlations and inter-orbital charge-charge correlations, so that they show nearly constant values from V to Fe: 1.57 (V and Cr), 1.52 (Mn), and 1.44 (Fe), which are roughly twice as large as those obtained by the d band model. The amplitudes of local moments are enhanced by the intra-orbital and inter-orbital spin-spin correlations and show large values for Mn and Fe: 2.87 (Mn) and 2.58 (Fe). These values are in good agreement with the experimental values estimated from the effective Bohr magneton number and the inner core photoemission data.

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

    International Nuclear Information System (INIS)

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

  2. Influence of rounds sub-grains in high burnup UO2 fuel

    International Nuclear Information System (INIS)

    In a recent paper, it has been shown that the so called 'rim structure' in high burnup UO2 fuel contains in fact two types of sub-grains: polyhedral and round ones. Polyhedral sub-grains have an average size of approximately 0.5 μm, and have been observed for more than ten years. The faceted porosity associated to these polyhedral sub-grains is characteristic of the rim effect. Round sub-grains have an average size of approximately 0.2 μm and are found to be formed on the free surface of initial grains or of polyhedral sub-grains. Round sub-grains can be observed in the rim area and also continuously from the periphery to the mid-pellet. This suggests that round sub-grains do not depend on rim effect, but are more likely to derive from a surface effect. In this contribution SEM photographs showing the evolution of the round sub-grains morphology will support a proposed mechanism for round sub-grains formation. This mechanism involves a surface modification due to stresses applied on a free surface using the Greenfeld formalism. These stresses could be due to segregation of fission products on some grain faces. This supposition is supported by EPMA experiments which show the segregation of some fission products on surfaces where round sub-grains are observed, while other surfaces with no round sub-grains have the same concentration in fission products as the bulk of the grains. Segregation of fission products on surface has also been observed in CANDU fuel by XPS. This specific behaviour of fission products gives a new insight in the chemistry of irradiated fuel and asks the question of the influence of round sub-grains formations on the release of fission products. (author)

  3. Use of axial burnup distribution profile in the nuclear safety analysis of spent nuclear fuel storage for WWER reactors in Ukraine

    International Nuclear Information System (INIS)

    The nuclear safety analysis of spent fuel storages taking into account fuel burnup should allow for burnup distribution along the height of the assembly. We propose a method based on an analysis of the axial burnup profiles of spent fuel assemblies. This method can be used in nuclear safety justification of spent fuel management and storage systems

  4. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    International Nuclear Information System (INIS)

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO2 fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  5. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

  6. Comparison between SERPENT and MONTEBURNS codes applied to burnup calculations of a GFR-like configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chersola, Davide [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Lomonaco, Guglielmo, E-mail: guglielmo.lomonaco@unige.it [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Marotta, Riccardo [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Mazzini, Guido [Centrum výzkumu Řež (Research Centre Rez), Husinec-Rez, cp. 130, 25068 Rez (Czech Republic)

    2014-07-01

    Highlights: • MC codes are widely adopted to analyze nuclear facilities, including GEN-IV reactors. • Burnup calculations are an efficient tool to test neutronic Monte Carlo codes. • In this comparison the used codes show some differences but a good agreement exists. - Abstract: This paper presents the comparison between two Monte Carlo based burnup codes: SERPENT and MONTEBURNS. Monte Carlo codes are fully and worldwide adopted to perform analyses on nuclear facilities, also in the frame of Generation IV advanced reactors simulations. Thus, faster and most powerful calculation codes are needed with the aim to analyze complex geometries and specific neutronic behaviors. Burnup calculations are an efficient tool to test neutronic Monte Carlo codes: indeed these calculations couple transport and depletion procedures, so that neutronic reactor behavior can be simulated in its totality. Comparisons have been performed on a configuration representing the Allegro MOX 75 MW{sub th} reactor proposed by the European GoFastR (Gas-cooled Fast Reactor) Project in the frame of the 7th Euratom Framework Program. Although in burnup and criticality comparisons the codes used in simulations show different calculation times and some differences in amounts and in precision (in term of statistical errors), a reasonably good agreement between them exists.

  7. Computational simulation of fuel burnup estimation for research reactors plate type

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadiasam@gmail.com [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research - International Atomic Energy Agency (MTR-IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx-Al) and a second composed with uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production. (author)

  8. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  9. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO2 fuel in Japan. The design concept should be compatible with UO2 fuel design. High burnup UO2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO2 8 x 8 array fuel developed for a second step of UO2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  10. Development of base technology for high burnup PWR fuel improvement Volume 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yang Eun; Lee, Sang Hee; Bae, Seong Man [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Chung, Jin Gon; Chung, Sun Kyo; Kim, Sun Du [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Kim, Jae Won; Chung, Sun Kyo; Kim, Sun Du [Korea Nuclear Fuel Development Inst., Seoul (Korea, Republic of)

    1995-12-31

    Development of base technology for high burnup nuclear fuel -Development of UO{sub 2} pellet manufacturing technology -Improvement of fuel rod performance code -Improvement of plenum spring design -Study on the mechanical characteristics of fuel cladding -Organization of fuel failure mechanism Establishment of next stage R and D program (author). 226 refs., 100 figs.

  11. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M. (Argonne National Lab., IL (USA))

    1989-09-01

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 {mu}m in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307{degree}C rather than the normal 288{degree}C, a relatively thick (50 to 70 {mu}m) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs.

  12. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-15

    Highlights: • Thermal properties of irradiated U–Mo alloy monolithic fuel samples were measured. • Density, thermal diffusivity, and thermal conductivity are influenced by increasing burnup. • U–Mo chemistry and specific heat capacity was not as sensitive to increasing burnup. • Thermal conductivity decreased approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} at 200 °C. • An empirical model developed previously agrees well with the experimental measurements. - Abstract: A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U–Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U–Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U–Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U–Mo alloy decreased approximately 30% for a fission density of 3.30 × 10{sup 21} fissions cm{sup −3} and approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  13. About a fuel for burnup reactor of periodical pulsed nuclear pumped laser

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, A.I.; Lukin, A.V.; Magda, L.E.; Magda, E.P.; Pogrebov, I.S.; Putnikov, I.S.; Khmelnitsky, D.V.; Scherbakov, A.P. [Russian Federal Nuclear Center, Snezhinsk (Russian Federation)

    1998-07-01

    A physical scheme of burnup reactor for a Periodic Pulsed Nuclear Pumped Laser was supposed. Calculations of its neutron physical parameters were made. The general layout and construction of basic elements of the reactor are discussed. The requirements for the fuel and fuel elements are established. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

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

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

    Science.gov (United States)

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

    2016-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2005-08-12

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

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

    Energy Technology Data Exchange (ETDEWEB)

    William J. Carmack

    2012-05-01

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

  18. Analytical and numerical study of radiation effect up to high burnup in power reactor fuels

    International Nuclear Information System (INIS)

    In the present work the behavior of fuel pellets for power reactors in the high burnup range (average burnup higher than 50 MWd/kgHM) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup, as long as a new microstructure develops, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behaviour. The evolution of porosity in the high burnup structure (HBS) is assumed to be determinant of the retention capacity of the fission gases released by the matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Starting from several works published in the open literature, a model was developed to describe the behaviour and evolution of porosity at local burnup values ranging from 60 to 300 MWd/KgHM. The model is mathematically expressed by a system of non-linear differential equations that take into account the open and closed porosity, the interactions between pores and the free surface and phenomena like pore's coalescence and migration and gas venting. Interactions of different orders between open and closed pores, growth of pores radius by vacancies trapping, the evolution of the pores number density, the internal pressure and over pressure within the pores, the fission gas retained in the matrix and released to the free volume are analyzed. The results of the simulations performed in the present work are in excellent agreement with experimental data available in the open literature and with results calculated by other authors (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2002-12-17

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-02-27

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

  1. Burnup determination and age dating of spent nuclear fuel using noble gas isotopic analysis

    International Nuclear Information System (INIS)

    During the chopping and dissolving phases of reprocessing, gases (such as tritium, krypton, xenon, iodine, carbon dioxide, nitrogen oxide, and steam) are released. These gases are traditionally transferred to a gas-treatment system for treatment, release, and/or recycle. Because of their chemically inert nature, the xenon and krypton noble gases are generally released directly into the loser atmosphere through the facility's stack. These gases (being fission products) contain information about the fuel being reprocessed and may prove a valuable monitor of reprocessing activities. Two properties of the fuel that may prove valuable from a safeguards standpoint are the fuel burnup and the fuel age (or time since discharge from the reactor). Both can be used to aid in confirming declared activities, and the burnup is generally indicative of the usability of the fuel for fabricating nuclear explosives. A study has been ongoing at Los Alamos National Laboratory to develop a methodology to determine spent-fuel parameters from measured xenon and/or krypton isotopic ratios on-stack at reprocessing facilities. This study has resulted in the generation of the NOVA data analysis code, which links to a comprehensive database of reactor physics parameters (calculated using the Monteburns 3.01 code system). NOVA has been satisfactorily tested for burnup determination of weapons-grade fuel from a US production reactor. Less effort has been spent quantifying NOVA's ability to predict burnup and fuel age for power reactor fuel. The authors describe the results predicted by NOVA for xenon and krypton isotopic ratios measured after the dissolution of spent-fuel samples from the Borssele reactor. The Borssele reactor is a 450-MW(electric) pressurized water reactor (PWR) consisting of 15 x 15 KWU assemblies. The spent-fuel samples analyzed were single fuel rods removed from one assembly and dissolved at the La Hague reprocessing facility. The assembly average burnup was estimated at 32

  2. Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)

    2008-04-15

    Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.

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

    International Nuclear Information System (INIS)

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

  4. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  5. Fission gas release behavior in high burnup UO2 fuels with developed rim-structure

    International Nuclear Information System (INIS)

    The effect of rim structure formation and external restraint pressure on fission gas release at transient conditions has been examined by using an out-of-pile high pressure heating technique for high burnup UO2 fuels (60, 74 and 90 GWd/t), which had been irradiated in test reactors. The latter two fuels bore a developed rim structure. The maximum heating temperature was 1500 degC, and the external pressures were independently controlled in the range of 10-150 MPa. The present high burnup fuel data were compared with those of previously studied BWR fuels of 37 and 54 GWd/t with almost no rim structure. The fission gas release and bubble swelling due to the growth of grain boundary bubbles and coarsened rim bubbles were effectively suppressed by the strong restraint pressure of 150 MPa for all the fuels; however the fission gas release remarkably increased for the two high burnup fuels with the developed rim structure, even at the strong restraint conditions. From the stepwise de-pressurization tests at an isothermal condition of 1500degC, the critical external pressure, below which a large burst release due to the rapid growth and interlinkage of the bubbles abruptly begins, was increased from a 40-60 MPa level for the middle burnup fuels to a high level of 120-140 MPa for the rim-structured high burnup fuels. The high potential for transient fission gas release and bubble swelling in the rim-structured fuels was attributed to highly over-pressurized fission gases in the rim bubbles. (author)

  6. Validation of the burn-up code EVOLCODE 2.0 with PWR experimental data and with a Sensitivity/Uncertainty analysis

    International Nuclear Information System (INIS)

    Highlights: • A successful validation of the burn-up simulation system EVOLCODE is presented here. • A Sensitivity/Uncertainty model was applied for uncertainty propagation/assessment. • Cross sections are for most cases the main contributors to inventory uncertainties. • The improved model helps to explain some simulation-experiment discrepancies. • Some hints for the improvement of basic data libraries are provided. - Abstract: A validation of the burn-up simulation system EVOLCODE 2.0 is presented here, involving the experimental measurement of U and Pu isotopes and some fission fragments production ratios after a burn-up of around 30 GWd/tU in a Pressurized Light Water Reactor (PWR). This work provides an in-depth analysis of the validation results, including the possible sources of the uncertainties. An uncertainty analysis based on the sensitivity methodology has been also performed, providing the uncertainties in the isotopic content propagated from the cross sections uncertainties. An improvement of the classical Sensitivity/Uncertainty (S/U) model has been developed to take into account the implicit dependence of the neutron flux normalization, that is, the effect of the constant power of the reactor. The improved S/U methodology, neglected in this kind of studies, has proven to be an important contribution to the explanation of some simulation-experiment discrepancies for which, in general, the cross section uncertainties are, for the most relevant actinides, an important contributor to the simulation uncertainties, of the same order of magnitude and sometimes even larger than the experimental uncertainties and the experiment-simulation differences. Additionally, some hints for the improvement of the JEFF3.1.1 fission yield library and for the correction of some errata in the experimental data are presented

  7. Experimental Fission Gas Release Determination at High Burnup by Means of Gamma Measurements on Fuel Rods in OL2

    International Nuclear Information System (INIS)

    hand, usually not so dominant with respect to the activity in the gas plenum. However, the longer cooling time the better quality of the spectrum due to the decay of the isotopes producing interfering gamma ray energies. The two fuel assemblies chosen for the measurement campaign in OL2 had been irradiated above the currently allowed maximum assembly burnup of 45 MWd/kgU. These two assemblies had average burnups of 50.4 and 50.9 MWd/kgU, respectively and were irradiated for four and five cycles, respectively. A total of 22 fuel rods from two fuel assemblies were selected for gamma measurements. These rods were chosen based on their calculated fission gas release, their location within the assembly, enrichment, Gadolinium content and rod diameter. The average burnup range for the measured rods was between 44 and 61 MWd/kgU. WSE has dedicated experimental equipment for the purpose of measuring the fission gas release, such as collimators and a detector system with software for data collecting and analysis. For these types of measurements, it is necessary to use a high resolution germanium detector due to the need to separate the three interfering peaks. At OL2, there is a location for a collimator built into the spent fuel pool wall that can be used, making it relatively simple to set up the detector system - as opposed to reactors without the built in collimator location which require that the detector be placed in a special submersible housing within the spent fuel pool. The built in collimator arrangement present at OL2 also features the possibility to change the characteristics of the collimator, if needed, such as the aperture size. For the measurements at OL2, a collimator of 20 by 20 millimeters was used to enhance the gamma ray intensity reaching the detector. Depending on the amount of fission gas present in the plenum volume of each rod, the needed measurement time can differ significantly from rod to rod. Due to the often limited time for measurements, a trade

  8. Advances in applications of burnup credit to enhance spent fuel transportation, storage, reprocessing and disposition. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    Given a trend towards higher burnup power reactor fuel, the IAEA began an active programme in burnup credit (BUC) with major meetings in 1997 (IAEA-TECDOC-1013), 2000 (IAEA-TECDOC-1241) and 2002 (IAEA-TECDOC-1378) exploring worldwide interest in using BUC in spent fuel management systems. This publication contains the proceedings of the IAEA's 4th major BUC meeting, held in London. Sixty participants from 18 countries addressed calculation methodology, validation and criticality, safety criteria, procedural compliance with safety criteria, benefits of BUC applications, and regulatory aspects in BUC. This meeting encouraged the IAEA to continue its activities on burnup credit including dissemination of related information, given the number of Member States having to deal with increased spent fuel quantities and extended durations. A 5th major meeting on burnup credit is planned 2008. Burnup credit is a concept that takes credit for the reduced reactivity of fuel discharged from the reactor to improve loading density of irradiated fuel assemblies in storage, transportation, and disposal applications, relative to the assumption of fresh fuel nuclide inventories in loading calculations. This report has described a general four phase approach to be considered in burnup credit implementation. Much if not all of the background research and data acquisition necessary for successful burnup credit development in preparation for licensing has been completed. Many fuel types, facilities, and analysis methods are encompassed in the public knowledge base, such that in many cases this guidance will provide a means for rapid development of a burnup credit program. For newer assembly designs, higher enrichment fuels, and more extensive nuclide credit, additional research and development may be necessary, but even this work can build on the foundation that has been established to date. Those, it is hoped that this report will serve as a starting point with sufficient reference to

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

    Science.gov (United States)

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

    1999-01-01

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

  10. Meta-Analysis with Complex Research Designs: Dealing with Dependence from Multiple Measures and Multiple Group Comparisons

    Science.gov (United States)

    Scammacca, Nancy; Roberts, Greg; Stuebing, Karla K.

    2014-01-01

    Previous research has shown that treating dependent effect sizes as independent inflates the variance of the mean effect size and introduces bias by giving studies with more effect sizes more weight in the meta-analysis. This article summarizes the different approaches to handling dependence that have been advocated by methodologists, some of…

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

    Science.gov (United States)

    Karpushkin, T. Yu.

    2012-12-01

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

  12. A Four Group Reference Code for Solving Neutron Diffusion Equation in a VVER-440 Core

    Energy Technology Data Exchange (ETDEWEB)

    Saarinen, Simo [Fortum Nuclear Services Ltd., P.O. Box 100, 00048 Fortum (Finland)

    2008-07-01

    Nuclear reactor core power calculation is essential in the analysis of the nuclear power plant and especially the core. Currently, the core power distribution in Loviisa VVER-440 core is calculated using nodal code HEXBU-3D and pin-power reconstruction code ELSI-1440 that solve the two group neutron diffusion equation. The computer power available has increased significantly during the last decades allowing us to develop a fine mesh code HEXRE for solving the four group diffusion equation. The diffusion equations are discretized using piecewise linear polynomials. The core is discretized using one node per fuel pin cell. The axial discretization can be chosen freely. The boundary conditions are described using diffusion theory and albedos. Burnup dependence is modelled by tabulating diffusion parameters at certain burnup values and using interpolation for the intermediate values. A two degree polynomial is used for the modelling of the feedback effects. Eigenvalue calculation for both boron concentration and multiplication factor control has been formulated. A possibility to perform fuel loading and shuffling operations is implemented. HEXRE has been thoroughly compared with HEXBU-3D and ELSI-1440. The effect of the different energy and space discretizations used is investigated. Some safety criteria for the core calculated with the HEXRE and HEXBU-3D/ELSI-1440 have been compared. From the calculations (e.g. the safety criteria) we can estimate whether there exists systematic deviations in HEXBU- 3D/ELSI-1440 calculations or not. (author)

  13. Classification of criticality calculations with correlation coefficient method and its application to OECD/NEA burnup credit benchmarks phase III-A and II-A

    International Nuclear Information System (INIS)

    A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the Nuclear Energy Agency of the Organisation for Economic Cooperation and Development (OECD/NEA). Phase III-A benchmark was a series of criticality calculations for irradiated Boiling Water Reactor (BWR) fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated Pressurized Water Reactor (PWR) fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark calculation results were classified according to the criterion that the values of the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of benchmark calculation results belonged to the same group, one calculation result was found predictable from the other. An example was shown for each of the Benchmarks. While the evaluated nuclear data seemed the main factor for the classification, further investigations were required for finding other factors. (author)

  14. Development of a Mobile CZT Detector System for Burnup Measurement of Spent Fuel Assembly and On-Site Application

    International Nuclear Information System (INIS)

    The advantages of mobile CdZnTe (CZT) detector for nuclear safeguard applications of spent fuel burnup inspection in assembly storage pond are compactness, low cost and ease of operations. In this work, a mobile detection system shield with tungsten alloy was designed and then performed on-site. Net count rate of the 662 keV line of 137Cs was produced linearly with burnup as experimental data simulations shows, in which the deviation from linearity is smaller than 9%. As a result, the feasibility of the method using CZT detector to monitor spent nuclear fuel assembly burnup in a fuel pond was validated. The results calculated with Monte Carlo procedure Geant4 can provide a theoretical guide for the further burnup measurement. (author)

  15. Validation of SWAT for burnup credit problems by analysis of post irradiation examination of 17*17 PWR fuel assembly

    International Nuclear Information System (INIS)

    For adopting burnup credit in transport or storage of spent fuel (SF), development of a reliable burnup calculation code is crucial. For this purpose, data of Post Irradiation Examination (PIE) have been extensively analyzed to evaluate accuracy of burnup calculation codes for a 14*14 or 15*15 PWR fuel assembly. This study shows results of analysis of this latest PIE with SWAT and ORIGEN2.1. SWAT is an integrated burnup code system for a 17*17 PWR fuel assembly that has been developed by Tohoku University and JAERI. The results show that SWAT can more precisely predict nuclide composition of latest PWR assembly than ORIGEN2.1. (O.M.)

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

    International Nuclear Information System (INIS)

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

  17. Hybrid Pareto artificial bee colony algorithm for multi-objective single machine group scheduling problem with sequence-dependent setup times and learning effects.

    Science.gov (United States)

    Yue, Lei; Guan, Zailin; Saif, Ullah; Zhang, Fei; Wang, Hao

    2016-01-01

    Group scheduling is significant for efficient and cost effective production system. However, there exist setup times between the groups, which require to decrease it by sequencing groups in an efficient way. Current research is focused on a sequence dependent group scheduling problem with an aim to minimize the makespan in addition to minimize the total weighted tardiness simultaneously. In most of the production scheduling problems, the processing time of jobs is assumed as fixed. However, the actual processing time of jobs may be reduced due to "learning effect". The integration of sequence dependent group scheduling problem with learning effects has been rarely considered in literature. Therefore, current research considers a single machine group scheduling problem with sequence dependent setup times and learning effects simultaneously. A novel hybrid Pareto artificial bee colony algorithm (HPABC) with some steps of genetic algorithm is proposed for current problem to get Pareto solutions. Furthermore, five different sizes of test problems (small, small medium, medium, large medium, large) are tested using proposed HPABC. Taguchi method is used to tune the effective parameters of the proposed HPABC for each problem category. The performance of HPABC is compared with three famous multi objective optimization algorithms, improved strength Pareto evolutionary algorithm (SPEA2), non-dominated sorting genetic algorithm II (NSGAII) and particle swarm optimization algorithm (PSO). Results indicate that HPABC outperforms SPEA2, NSGAII and PSO and gives better Pareto optimal solutions in terms of diversity and quality for almost all the instances of the different sizes of problems.

  18. Hybrid Pareto artificial bee colony algorithm for multi-objective single machine group scheduling problem with sequence-dependent setup times and learning effects.

    Science.gov (United States)

    Yue, Lei; Guan, Zailin; Saif, Ullah; Zhang, Fei; Wang, Hao

    2016-01-01

    Group scheduling is significant for efficient and cost effective production system. However, there exist setup times between the groups, which require to decrease it by sequencing groups in an efficient way. Current research is focused on a sequence dependent group scheduling problem with an aim to minimize the makespan in addition to minimize the total weighted tardiness simultaneously. In most of the production scheduling problems, the processing time of jobs is assumed as fixed. However, the actual processing time of jobs may be reduced due to "learning effect". The integration of sequence dependent group scheduling problem with learning effects has been rarely considered in literature. Therefore, current research considers a single machine group scheduling problem with sequence dependent setup times and learning effects simultaneously. A novel hybrid Pareto artificial bee colony algorithm (HPABC) with some steps of genetic algorithm is proposed for current problem to get Pareto solutions. Furthermore, five different sizes of test problems (small, small medium, medium, large medium, large) are tested using proposed HPABC. Taguchi method is used to tune the effective parameters of the proposed HPABC for each problem category. The performance of HPABC is compared with three famous multi objective optimization algorithms, improved strength Pareto evolutionary algorithm (SPEA2), non-dominated sorting genetic algorithm II (NSGAII) and particle swarm optimization algorithm (PSO). Results indicate that HPABC outperforms SPEA2, NSGAII and PSO and gives better Pareto optimal solutions in terms of diversity and quality for almost all the instances of the different sizes of problems. PMID:27652166

  19. Structure-dependent tuning of electro-optic and thermoplastic properties in triphenyl groups containing molecular glasses

    Energy Technology Data Exchange (ETDEWEB)

    Traskovskis, Kaspars, E-mail: kaspars.traskovskis@rtu.lv [Riga Technical University, Faculty of Materials Science and Applied Chemistry, 3/7 Paula Valdena Street, Riga LV-1048 (Latvia); Zarins, Elmars; Laipniece, Lauma [Riga Technical University, Faculty of Materials Science and Applied Chemistry, 3/7 Paula Valdena Street, Riga LV-1048 (Latvia); Tokmakovs, Andrejs [Institute of Solid State Physics, University of Latvia, 8 Kengaraga Street, Riga LV-1063 (Latvia); Kokars, Valdis [Riga Technical University, Faculty of Materials Science and Applied Chemistry, 3/7 Paula Valdena Street, Riga LV-1048 (Latvia); Rutkis, Martins [Institute of Solid State Physics, University of Latvia, 8 Kengaraga Street, Riga LV-1063 (Latvia)

    2015-04-01

    The series of seven molecular compounds composed of D–π–A chromophores and triphenylmethyl auxiliary groups were characterized by UV–Vis spectroscopy, differential scanning calorimetry and quantum chemical calculations. Nonlinear optical (NLO) properties of compounds were determined by second harmonic generation measurements in corona poled thin glassy films. The results show that triphenylmethyl auxiliary groups are effective at shielding undesirable dipole interactions in solid phase thus increasing NLO efficiency of materials. Thermal stability up to 108 °C was achieved for a polar order in poled samples. - Highlights: • Triphenylmethyl groups can be used to reduce solid phase dipole interactions in organic molecular materials. • NLO efficiency of a poled material is higher, if a number of present triphenyl groups increases. • NLO efficiency of materials decreases, if polarity of used chromophores increases. • Thermal stability of polar order up to 108 °C can be achieved in poled organic glasses.

  20. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel

    International Nuclear Information System (INIS)

    isotope concentrations in fission gas inclusions have not been reported yet. The aim of this work was focused on quantitative analysis of Xe isotope concentrations in inclusions of high burn-up samples to determine gas pressures within fuel pores. Therefore, various calibration strategies were studied in great detail. The investigation of Xe-implantation showed that the uncertainties in Xe ion implantation are too high to produce suitable Xe-implanted standard materials. The generation of Xe-bubbles in Plexiglass was also investigated and Xe inclusions of different bubble sizes were generated. However, the size determinations of the inclusions, the opening and possible gas losses through the Plexiglass were factors preventing accurate Xe quantification. The use of direct Xe gas injection was considered as a suitable strategy for establishing calibration curves. The Xe response was studied in dependence on gas concentration, gas volume and presence or absence of matrix in the ICP (Inductively Coupled Plasma). To investigate the ionization behavior of Xe in the ICP, the ablation behavior of ZrO2 and UO2 was determined in dependence on different wavelengths (266 nm, 193 nm), crater diameters, repetition rates and carrier gases (Ar, He). The mass ablated per laser pulse was very similar for the two matrices. The simultaneous introduction of matrix and gas into the ICP showed insignificant matrix effects. Even the absence of matrix effects was shown, filtering of the matrix before entering the plasma was evaluated to be crucial. Applying a filter leads to a complete gas/solid separation and opens access to the matrix interfering 134Xe and 136Xe isotopes. Thus, calibration gas addition in combination with an aerosol filter was the key strategy for quantification of Xe fission gas production in high burn-up fuel samples. Gas addition in volumes of 1 to 100 μl and at a concentration of 20 and 200 μg/g provided linear calibration curves with a regression coefficient of 0

  1. Derivation of a stable coupling scheme for Monte Carlo burnup calculations with the thermal-hydraulic feedback

    OpenAIRE

    Dufek, Jan; Anglart, Henryk

    2013-01-01

    Numerically stable Monte Carlo burnup calculations of nuclear fuel cycles are now possible with the previously derived Stochastic Implicit Euler method based coupling scheme. In this paper, we show that this scheme can be easily extended to include the thermal-hydraulic feedback during the Monte Carlo burnup simulations, while preserving its unconditional stability property. At each time step, the implicit solution (for the end-of-step neutron flux, fuel nuclide densities and thermal-hydrauli...

  2. Using SERPENT Monte Carlo and Burnup code to model Traveling Wave Reactors - TWR

    International Nuclear Information System (INIS)

    This paper is mainly devoted to the proof-of-principle implementation of the SERPENT code for the simulation of traveling wave reactors. Traveling wave reactors are both fast reactors and nuclear burning wave reactors in which the breeding and burning of nuclear fuel appear almost simultaneously. SERPENT is a neutron transport code whose last official update package is SERPENT 1.1.19 and whose SERPENT 2 version is currently in progress. The investigation of SERPENT 1.1.19 and of SERPENT 2 codes for multiprocessor tasks with long burnup steps was performed. It appears that SERPENT 2 has eliminated parallelization problems efficiently. Methods to remove the influence of the ignition zone were considered, and neutron transport simulations with various fragmentations of the burnup zone were performed. (authors)

  3. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    Energy Technology Data Exchange (ETDEWEB)

    Kee, E. [Houston Lighting & Power Co., Wadworth, TX (United States)

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  4. Draft evaluation of the frequency for gas sampling for the high burnup confirmatory data project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-26

    This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.

  5. Computation of classical triton burnup with high plasma temperature and current

    International Nuclear Information System (INIS)

    For comparison with experiment, the expected production of 14-MeV neutrons from the burnup of tritons produced in the d(d,t)p reaction must be computed. An effort was undertaken to compare in detail the computer codes used for this purpose at TFTR and JET. The calculation of the confined fraction of tritons by the different codes agrees to within a few percent. The high electron temperature in the experiments has raised the critical energy of the tritons that are slowing down to near or above the peak of the D-T reactivity, making the ion drag terms more important. When the different codes use the same slowing down formulas, the calculated burnup was within 6% for a case where orbit effects are expected to be small. Then results from codes with and without the effects of finite radial orbit excursions were compared for two test cases. For medium to high current discharges the finite radius effects are only of order 10%. A new version of the TFTR burnup code using an implicit Fokker-Planck solution was written to include the effects of energy diffusion and charge exchange. These effects change the time-integrated yields by only a few percent, but can significantly affect the instantaneous rates in time. Significant populations of hot ions can affect the fusion reactivity, and this effect was also studied. In particular, the d(d,p)t rate can be 10%--15% less than the d(d,3He)n rate which is usually used as a direct monitor of the triton source. Finally, a finite particle confinement time for the thermalized tritons can increase the apparent ''burn-up'' either if there is a high thermal deuteron temperature or if there exists a significant beam deuteron density

  6. Estimate of preliminary experiments to study the burn-up of gadolinium as a poison

    International Nuclear Information System (INIS)

    Full text: Proposed preliminary experiments to determine the burn-up of Gd2O3 as a poison in different reactors are discussed. Estimates are given of parameters such as the weight of the sample to be irradiated, irradiation and decay times, expected activity and photon spectrum. 1 g samples of natural UO2 with 8 % of Gd2O3, 3 days irradiation time and 30 days decay time are recommended

  7. Angra 1 high burnup fuel behaviour under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: dsgomes@ipen.b, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The 16x16 NGF (Next Generation Fuel) fuel assembly, comprising of highly corrosive-resistant ZIRLO clad fuel rods, been replacing the current 16x16 Standard (16STD) fuel assembly in the Angra 1, a pressurized water reactor, with a net output of 626 MWe. The 16x16 NGF fuel assemblies are designed for a peak rod average burnup of up to 75 GWd/MTU, thus improving fuel utilization and reducing spent fuel storage issues. A design basis accident, the Reactivity Initiated Accident (RIA), became a concern for a further increase in burnup as the simulated RIA tests revealed a lower enthalpy threshold for fuel failure. Two fuel performance codes, FRAPCON and FRAPTRAN, were used to predict high burnup behavior of Angra 1, during an RIA. The maximum average linear fuel rating used was 17.62 KW/m. The FRAPCON 3.4 code was applied to simulate the steady-state performance of the 16 NGF fuel rods up to a burnup of 55 GWd/MTU. With FRAPTRAN-1.4 the fuel behavior was simulated for an RIA power pulse of 4.5 ms (FHWH), and enthalpy peak of 130 Cal/g. With FRAPCON-3.4, the corrosion and hydrogen pickup characteristics of the advanced ZIRLO clad fuel rods were added to the code by modifying the actual corrosion model for Zircaloy-4 through the multiplication of empirical factors, which were appropriate to each alloy, and by means of reducing the current hydrogen pickup fraction. (author)

  8. A Genesis breakup and burnup analysis in off-nominal Earth return and atmospheric entry

    Science.gov (United States)

    Salama, Ahmed; Ling, Lisa; McRonald, Angus

    2005-01-01

    The Genesis project conducted a detailed breakup/burnup analysis before the Earth return to determine if any spacecraft component could survive and reach the ground intact in case of an off-nominal entry. In addition, an independent JPL team was chartered with the responsibility of analyzing several definitive breakup scenarios to verify the official project analysis. This paper presents the analysis and results of this independent team.

  9. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

    Energy Technology Data Exchange (ETDEWEB)

    Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

    2006-10-31

    The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty,cycles (high burnup, boiling, aggressive chemistry) andto investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment.

  10. Evaluation of safety criteria on LOCA and RIA for high burnup nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Ki; Bang, Je Geon; Kim, Dae Ho; Yang, Yong Sik; Song, Keun Woo

    2007-12-15

    Comprehensive researches in many countries and some international research programs to investigate the applicability to high burnup nuclear fuels have been performing as the existing safety criteria of DBA such as LOCA and RIA was established several decades ago. In this report, main research programs for the safety criteria of DBA such as LOCA and RIA are introduced, and also the current status on the modification of the safety criteria are also introduced.

  11. Evaluation of safety criteria on LOCA and RIA for high burnup nuclear fuel

    International Nuclear Information System (INIS)

    Comprehensive researches in many countries and some international research programs to investigate the applicability to high burnup nuclear fuels have been performing as the existing safety criteria of DBA such as LOCA and RIA was established several decades ago. In this report, main research programs for the safety criteria of DBA such as LOCA and RIA are introduced, and also the current status on the modification of the safety criteria are also introduced

  12. Development of the CANDU high-burnup fuel design/analysis technology

    International Nuclear Information System (INIS)

    This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs

  13. Development of a fuel rod thermal-mechanical analysis code for high burnup fuel

    International Nuclear Information System (INIS)

    The thermal-mechanical analysis code for high burnup BWR fuel rod has been developed by NFI. The irradiation data accumulated up to the assembly burnup of 55 GWd/t in commercial BWRs were adopted for the modeling. In the code, pellet thermal conductivity degradation with burnup progress was considered. Effects of the soluble FPs, irradiation defects and porosity increase due to RIM effect were taken into the model. In addition to the pellet thermal conductivity degradation, the pellet swelling due to the RIM porosity was studied. The modeling for the high burnup effects was also carried out for (U, Gd)O2 and MOX fuel. The thermal conductivities of all pellet types, UO2, (U, Gd)O2 and (U, Pu)O2 pellets, are expressed by the same form of equation with individual coefficient γ in the code. The pellet center temperature was calculated using this modeling code, and compared with measured values for the code verification. The pellet center temperature calculated using the thermal conductivity degradation model agreed well with the measured values within ±150 deg. C. The influence of rim porosity on pellet center temperature is small, and the temperature increase in only 30 deg. C at 75 GWd/t and 200 W/cm. The pellet center temperature of MOX fuel was also calculated, and it was found that the pellet center temperature of MOX fuel with 10wt% PuO2 is about 60 deg. C higher than UO2 fuel at 75 GWd/t and 200 W/cm. (author)

  14. Details on an actinide-only burnup credit application in the USA

    International Nuclear Information System (INIS)

    Details on the Actinide-Only burnup credit assumptions that will be used for the CASTOR X/32 S cask are presented. Preliminary results show that using a conservative set of assumptions the cask will allow most fuel to be loaded without the addition of any additional reactivity control. With the addition of 8 control rod elements it is possible to load most of the rest of the fuel. (author)

  15. Prototype studies on the nondestructive online burnup determination for the modular pebble bed reactors

    International Nuclear Information System (INIS)

    Highlights: • Prototype study of online burnup measurement for HTR proves its feasibility. • Calibration and its correction of burnup assay device is discussed and verified. • Analysis of simulated gamma spectra shows good performance of spectra-unfolding method. - Abstract: The online fuel pebble burnup determination in future modular pebble bed reactor is implemented by measuring nondestructively the activity of the monitoring nuclide Cs-137 with HPGe detector on a pebble-by-pebble basis. Based on a full size prototype the feasibility is investigated. The prototype was first tested by using double sources to show that a precision of 2.8% (1σ) can be achieved in the determination of the Cs-137 net counting rate. Then, the relationship between the Cs-137 activity and the net counting rate recorded in the HPGe detector is calibrated with a standard Cs-137 source contained in the center of a graphite sphere with the same dimension as a real fuel pebble. Because the self attenuation of the calibration source differs with a fuel pebble, a correction factor of 1.07 ± 0.02 (p = 0.95) to the calibration is derived by using the efficiency transfer method. Last, by analyzing the spectra generated with KORIGEN software followed by Monte Carlo simulation, it is predicted that the relative standard deviation of the Cs-137 net counting rate can be still controlled below 3.5% despite of the presence of all the interfering peaks. The results demonstrate the feasibility of utilizing HPGe gamma spectrometry in the online determination of the pebble burnup in future modular pebble bed reactors

  16. Investigation of burnup credit allowance in the criticality safety evaluation of spent fuel casks

    International Nuclear Information System (INIS)

    This presentation discusses work in progress on criticality analysis verification for designs which take account of the burnup and age of transported fuel. The work includes verification of cross section data, correlation with experiments, proper extension of the methods into regimes not covered by experiments, establishing adequate reactivity margins, and complete documentation of the project. Recommendations for safe operational procedures are included, as well as a discussion of the economic and safety benefits of such designs

  17. Investigation of research and development subjects for the Very High Burnup Fuel. Development of fuel pellet

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Kimio; Amano, Hidetoshi; Suzuki, Yasufumi; Furuta, Teruo; Nagase, Fumihisa; Suzuki, Masahide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1993-06-01

    A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author).

  18. Investigation of research and development subjects for the Very High Burnup Fuel

    International Nuclear Information System (INIS)

    A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author)

  19. The formation process of the pellet-cladding bonding layer in high burnup BWR fuels

    International Nuclear Information System (INIS)

    The bonding formation process was studied by EPMA analysis, XRD measurements, and SEM/TEM observations for the oxide layer on a cladding inner surface and the pellet-cladding bonding layer in irradiated fuel rods. Specimens were prepared from fuels which had been irradiated to the pellet average burnups of 15, 27, 42 and 49 GWd/t in BWRs. In the lower burnup specimens of 15 and 27 GWd/t, no bonding layer was found, while the higher burnup specimens of 42 and 49 GWd/t had a typical bonding layer about 10 to 20 μm thick. A bonding layer which consisted of two regions was found in the latter fuels. One region of the inner surface of the Zr liner cladding was made up mainly of ZrO2 with a small amount of dissolved UO2. The structure of this ZrO2 consisted of cubic polycrystals a few nanometers in size, while no monoclinic crystals were found. The other region, near the pellet surface, had both a cubic solid solution of (U,Zr)O2 and amorphous phase in which the concentrations of UO2 and ZrO2 changed continuously. Even in the lower burnup specimens having no bonding layer, cubic ZrO2 phase was identified in the cladding inner oxide layer. The XRD measurements were consistent with the TEM results of the absence of the monoclinic ZrO2 phase. Phase transformation and amorphization were attributed to fission damage, since such phenomena have never been observed in the cladding outer surface. Phase transformation from monoclinic to cubic ZrO2 and amorphization by irradiation damage of fission products were discussed in connection with the formation mechanism and conditions of the bonding layer. (author)

  20. Burn-up cross sections of 51Cr, 59Fe, 65Zn, 86Rb, 103Ru

    International Nuclear Information System (INIS)

    Targets of Cr, Fe, Zn, Rb, and Ru were irradiated in the hydraulic tube of the Oak Ridge HFIR reactor at a neutron flux of 2.6 x 1015 n/cm2sec for 1 day and 20 days. The reactor burn-up cross sections (in barns) of the radioactive product nuclides are: 51Cr, 59Fe, 65Zn, 60 +- 30; 86Rb, 103Ru, <20

  1. Assessing the Effect of Fuel Burnup on Control Rod Worth for HEU and LEU Cores of Gharr-1

    Directory of Open Access Journals (Sweden)

    E.K. Boafo

    2013-02-01

    Full Text Available An important parameter in the design and analysis of a nuclear reactor is the reactivity worth of the control rod which is a measure of the efficiency of the control rod to absorb excess reactivity. During reactor operation, the control rod worth is affected by factors such as the fuel burnup, Xenon concentration, Samarium concentration and the position of the control rod in the core. This study investigates the effect of fuel burnup on the control rod worth by comparing results of a fresh and an irradiated core of Ghana's Miniature Neutron Source Reactor for both HEU and LEU cores. In this study, two codes have been utilized namely BURNPRO for fuel burnup calculation and MCNP5 which uses densities of actinides of the irradiated fuel obtained from BURNPRO. Results showed a decrease of the control rod worth with burnup for the LEU while rod worth increased with burnup for the HEU core. The average thermal flux in both inner and outer irradiation sites also decreased significantly with burnup for both cores.

  2. Characterisation of high-burnup LWR fuel rods through gamma tomography

    International Nuclear Information System (INIS)

    Current fuel management strategies for light water reactors (LWRs), in countries with high back-end costs, progressively extend the discharge burnup at the expense of increasing the 235U enrichment of the fresh UO2 fuel loaded. In this perspective, standard non-destructive assay techniques, which are very attractive because they are fast, cheap, and preserve the fuel integrity, in contrast to destructive approaches, require further validation when burnup values become higher than 50 GWd/t. This doctoral work has been devoted to the development and optimisation of non-destructive assay techniques based on gamma-ray emissions from irradiated fuel. It represents an important extension of the unique, high-burnup related database, generated in the framework of the LWR PROTEUS Phase II experiments. A novel tomographic measurement station has been designed and developed for the investigation of irradiated fuel rod segments. A unique feature of the station is that it allows both gamma-ray transmission and emission computerised tomography to be performed on single fuel rods. Four burnt UO2 fuel rod segments of 400 mm length have been investigated, two with very high (52 GWd/t and 71 GWd/t) and two with ultra-high (91 GWd/t and 126 GWd/t) burnup. Several research areas have been addressed, as described below. The application of transmission tomography to spent fuel rods has been a major task, because of difficulties of implementation and the uniqueness of the experiments. The main achievements, in this context, have been the determination of fuel rod average material density (a linear relationship between density and burnup was established), fuel rod linear attenuation coefficient distribution (for use in emission tomography), and fuel rod material density distribution. The non-destructive technique of emission computerised tomography (CT) has been applied to the very high and ultra-high burnup fuel rod samples for determining their within-rod distributions of caesium and

  3. Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Venkiteswaran, C.N., E-mail: cnv@igcar.gov.in; Jayaraj, V.V.; Ojha, B.K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B.P.C.; Kasiviswanathan, K.V.; Jayakumar, T.

    2014-06-01

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

  4. Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library

    Directory of Open Access Journals (Sweden)

    Lecarpentier D.

    2013-03-01

    Full Text Available Burnup Credit (BUC is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a “best estimate” value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 FPs of PWRMOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF-3.1.1/SHEM library. Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections.

  5. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  6. Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Parks, C V; DeHart, M D [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wagner, John C [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2000-03-13

    This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan.

  7. Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel

    International Nuclear Information System (INIS)

    This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan

  8. Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library

    International Nuclear Information System (INIS)

    Burnup Credit (BUC) is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a 'best estimate' value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 fission products (FPs) of PWR-MOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF- 3.1.1/SHEM library). Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections. (authors)

  9. OECD/NEA burnup credit criticality benchmark. Result of phase IIA

    International Nuclear Information System (INIS)

    The report describes the final result of the Phase IIA of the Burnup Credit Criticality Benchmark conducted by OECD/NEA. In the Phase IIA benchmark problems, the effect of an axial burnup profile of PWR spent fuels on criticality (end effect) has been studied. The axial profiles at 10, 30 and 50 GWd/t burnup have been considered. In total, 22 results from 18 institutes of 10 countries have been submitted. The calculated multiplication factors from the participants have lain within the band of ± 1% Δk. For the irradiation up to 30 GWd/t, the end effect has been found to be less than 1.0% Δk. But, for the 50 GWd/t case, the effect is more than 4.0% Δk when both actinides and FPs are taken into account, whereas it remains less than 1.0% Δk when only actinides are considered. The fission density data have indicated the importance end regions have in the criticality safety analysis of spent fuel systems. (author)

  10. Evaluation of the characteristics of high burnup and high plutonium content mixed oxide (MOX) fuel

    International Nuclear Information System (INIS)

    Two kinds of MOX fuel irradiation tests, i.e., MOX irradiation test up to high burnup and MOX having high plutonium content irradiation test, have been performed from JFY 2007 for five years in order to establish technical data concerning MOX fuel behavior during irradiation, which shall be needed in safety regulation of MOX fuel with high reliability. The high burnup MOX irradiation test consists of irradiation extension and post irradiation examination (PIE). The activities done in JFY 2011 are destructive post irradiation examination (D-PIE) such as EPMA and SIMS at CEA (Commissariat a l'Enegie Atomique) facility. Cadarache and PIE data analysis. In the frame of irradiation test of high plutonium content MOX fuel programme, MOX fuel rods with about 14wt % Pu content are being irradiated at BR-2 reactor and corresponding PIE is also being done at PIE facility (SCK/CEN: Studiecentrum voor Kernenergie/Centre d'Etude l'Energie Nucleaire) in Belgium. The activities done in JFY 2011 are non-destructive post irradiation examination (ND-PIE) and D-PIE and PIE data analysis. In this report the results of EPMA and SIMS with high burnup irradiation test and the result of gamma spectrometry measurement which can give FP gas release rate are reported. (author)

  11. Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Asah-Opoku Fiifi

    2014-01-01

    Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

  12. Neurocognitive Dysfunction in Bipolar and Schizophrenia Spectrum Disorders Depends on History of Psychosis Rather Than Diagnostic Group

    OpenAIRE

    Simonsen, Carmen; Sundet, Kjetil; Vaskinn, Anja; Birkenaes, Astrid B.; Engh, John A; Færden, Ann; Jónsdóttir, Halldóra; Ringen, Petter Andreas; Opjordsmoen, Stein; Melle, Ingrid; Friis, Svein; Andreassen, Ole A

    2009-01-01

    Objectives: Neurocognitive dysfunction is milder in bipolar disorders than in schizophrenia spectrum disorders, supporting a dimensional approach to severe mental disorders. The aim of this study was to investigate the role of lifetime history of psychosis for neurocognitive functioning across these disorders. We asked whether neurocognitive dysfunction in bipolar and schizophrenia spectrum disorders depends more on history of psychosis than diagnostic category or subtype. Methods: A sample o...

  13. An empirical formulation to describe the evolution of the high burnup structure

    Science.gov (United States)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-01

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  14. An empirical formulation to describe the evolution of the high burnup structure

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-15

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  15. Ion mobility studies of carbohydrates as group I adducts: isomer specific collisional cross section dependence on metal ion radius.

    Science.gov (United States)

    Huang, Yuting; Dodds, Eric D

    2013-10-15

    Carbohydrates play numerous critical roles in biological systems. Characterization of oligosaccharide structures is essential to a complete understanding of their functions in biological processes; nevertheless, their structural determination remains challenging in part due to isomerism. Ion mobility spectrometry provides the means to resolve gas phase ions on the basis of their shape-to-charge ratios, thus providing significant potential for separation and differentiation of carbohydrate isomers. Here, we report on the determination of collisional cross sections for four groups of isomeric carbohydrates (including five isomeric disaccharides, four isomeric trisaccharides, two isomeric pentasaccharides, and two isomeric hexasaccharides) as their group I metal ion adducts (i.e., [M + Li](+), [M + Na](+), [M + K](+), [M + Rb](+), and [M + Cs](+)). In all, 65 collisional cross sections were measured, the great majority of which have not been previously reported. As anticipated, the collisional cross sections of the carbohydrate metal ion adducts generally increase with increasing metal ion radius; however, the collisional cross sections were found to scale with the group I cation size in isomer specific manners. Such measurements are of substantial analytical value, as they illustrate how the selection of charge carrier influences carbohydrate ion mobility determinations. For example, certain pairs of isomeric carbohydrates assume unique collisional cross sections upon binding one metal ion, but not another. On the whole, these data suggest a role for the charge carrier as a probe of carbohydrate structure and thus have significant implications for the continued development and application of ion mobility spectrometry for the distinction and resolution of isomeric carbohydrates.

  16. Effects of the Scale-dependent Vacuum Expectation Values in the Renormalization Group Analysis of Neutrino Masses

    CERN Document Server

    Singh, N N

    2001-01-01

    The contribution of scale-dependent vacuum expectation values (VEVs) of Higgs scalars, which gives significant effects in the evolution of fundamental fermion masses in the Minimal Supersymmetric Standard Model (MSSM), is now considered in the derivation of the one-loop analytic expression for the evolution of the left-handed Majorana neutrino masses with energies. The inclusion of such effect of the running VEV increases the stability of the neutrino masses under quantum corrections and leads to a mild decrease of neutrino masses with higher energies.

  17. TOWARDS LEARNER-CENTRED MEDICAL CURRICULUM: QUALITATIVE FOCUS GROUP STUDY OF INDIVIDUAL PREFERENCES DEPENDING ON VERBAL ENVIRONMENT.

    Science.gov (United States)

    Bukeyeva, A; Riklefs, V; Riklefs, I; Tashkenbayeva, V; Kassatova, A

    2016-05-01

    There is a strong evidence in medical education literature that the learner-centred curriculum favouring the use of metacognition and self-learning is very proficient. However, ethnocultural and verbal environment may undermine learners' ability to utilise the learning strategies, leading to inefficient learning. This study aimed to investigate the personal preferences of learners in multilingual educational environment prompting the most efficient learning. The study uses qualitative focus group methodology to understand students' opinion on how educational environment influences the efficiency of medical school curriculum. PMID:27348171

  18. Risk group dependence of dose-response for biopsy outcome after three-dimensional conformal radiation therapy of prostate cancer

    International Nuclear Information System (INIS)

    Background and purpose: We fit phenomenological tumor control probability (TCP) models to biopsy outcome after three-dimensional conformal radiation therapy (3D-CRT) of prostate cancer patients to quantify the local dose-response of prostate cancer. Materials and methods: We analyzed the outcome after photon beam 3D-CRT of 103 patients with stage T1c-T3 prostate cancer treated at Memorial Sloan-Kettering Cancer Center (MSKCC) (prescribed target doses between 64.8 and 81 Gy) who had a prostate biopsy performed ≥2.5 years after end of treatment. A univariate logistic regression model based on Dmean (mean dose in the planning target volume of each patient) was fit to the whole data set and separately to subgroups characterized by low and high values of tumor-related prognostic factors T-stage (6), and pre-treatment prostate-specific antigen (PSA) (≤10 ng/ml vs. >10 ng/ml). In addition, we evaluated five different classifications of the patients into three risk groups, based on all possible combinations of two or three prognostic factors, and fit bivariate logistic regression models with Dmean and the risk group category to all patients. Dose-response curves were characterized by TCD50, the dose to control 50% of the tumors, and γ50, the normalized slope of the dose-response curve at TCD50. Results: Dmean correlates significantly with biopsy outcome in all patient subgroups and larger values of TCD50 are observed for patients with unfavorable compared to favorable prognostic factors. For example, TCD50 for high T-stage patients is 7 Gy higher than for low T-stage patients. For all evaluated risk group definitions, Dmean and the risk group category are independent predictors of biopsy outcome in bivariate analysis. The fit values of TCD50 show a clear separation of 9-10.6 Gy between low and high risk patients. The corresponding dose-response curves are steeper (γ50=3.4-5.2) than those obtained when all patients are analyzed together (γ50=2.9). Conclusions: Dose

  19. Mortality rates of 0-group plaice ( Platessa platessa L.), dab ( Limanda limanda L.) and turbot ( Scophthalmus maximus L.) in European waters . III. Density dependence of mortality rates of 0-group plaice and some demographic implications

    Science.gov (United States)

    Beverton, R. J. H.; Iles, T. C.

    This last of our three linked contributions on the dynamics of North Sea plaice is concerned with the detection and measurement from demographic data of the density-dependence of mortality rate during the early demersal phase of the life history. A mathematical expression is developed for the survival trajectory of a cohort subject to an instantaneous relative mortality rate which is a linear function of the algorithm of its density. This is used to analyse three independent data sets; a. pairs of densities at or soon after settlement in the Wadden Sea, b. estimates of seasonal mortality rates (M̊ d -1) and initial density of 0-group plaice cohorts derived by linear regression and c. autumn estimates 0- and 1-group fish from the ICES Young Fish Surveys. After correcting for various sources of bias, these each gave statistically significant estimates of the density-dependent mortality coefficient μ2 of 0.015, 0.0044 and 0.0010 per day, respectively. The same theoretical treatment of density-dependent mortality is used to develop an equation predicting the progressive 'damping' of the extremes of inter-year-class variation with age. The above three estimates of the density-dependent mortality coefficient μ2 applied in sequence provide more than sufficient 'damping' to explain the very low variability of recruitment and long-term stability which is characteristic of the North Sea plaice stock.

  20. Effects of the scale-dependent vacuum expectation values in the renormalisation group analysis of neutrino masses

    Energy Technology Data Exchange (ETDEWEB)

    Singh, N.N. [Dept. of Physics and Astronomy, Univ. of Southampton (United Kingdom)

    2001-02-01

    The contribution of scale-dependent vacuum expectation values (VEVs) of Higgs scalars, which gives significant effects in the evolution of the fundamental fermion masses in the minimal supersymmetric standard model (MSSM), is now considered in the derivation of the analytic one-loop expression for the evolution of the left-handed Majorana neutrino masses with energies. The inclusion of such an effect of the running VEV increases the stability of the neutrino masses under quantum corrections even for the low values of tan {beta}{>=}1.42 at the scale {mu}=10{sup 12} GeV, and leads to a mild decrease of the neutrino masses with higher energies. Such a trend is common with that of other fundamental fermion masses. (orig.)

  1. Effects of the scale-dependent vacuum expectation values in the renormalisation group analysis of neutrino masses

    International Nuclear Information System (INIS)

    The contribution of scale-dependent vacuum expectation values (VEVs) of Higgs scalars, which gives significant effects in the evolution of the fundamental fermion masses in the minimal supersymmetric standard model (MSSM), is now considered in the derivation of the analytic one-loop expression for the evolution of the left-handed Majorana neutrino masses with energies. The inclusion of such an effect of the running VEV increases the stability of the neutrino masses under quantum corrections even for the low values of tan β≥1.42 at the scale μ=1012 GeV, and leads to a mild decrease of the neutrino masses with higher energies. Such a trend is common with that of other fundamental fermion masses. (orig.)

  2. Efficacy of treatment in an opioid –dependent population group using the Maudsley Addiction Profile (MAP) tool

    Science.gov (United States)

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

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-11

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

  4. The impact of phenotypic appearance on body weight and egg production in laying hens: a group-size- and experience-dependent phenomenon.

    Science.gov (United States)

    Marin, R H; Liste, M G; Campderrich, I; Estevez, I

    2014-07-01

    Alterations of birds' phenotypic appearance (PA) may lead to unwanted behaviors, potentially impairing poultry welfare, health, and productive performance. Likewise, group size may play an important role modulating the expression of adaptive behaviors. This study evaluates whether changes in the PA of Hy-line Brown laying hens may affect their BW and egg production, and if so, whether these effects depend on group size. A total of 1,050 one-day-old chicks were randomly assigned to 1 of 45 pens. Groups were of 10, 20, or 40 individuals (8 hens/m(2)). At arrival, the PA of 0, 30, 50, 70, or 100% of the birds within each group was artificially altered by marking the back of their heads black. The remaining birds within groups were unaltered. The 30% marked hens within groups of 10 individuals had a lower BW at 24 wk of age than their 70% unmarked counterparts, whereas the other groups showed similar BW. No differences were detected in egg laying performance during this phase. Next, within the initially homogeneous groups (0 and 100%), 30, 50, and 70% of the hens were either marked or unmarked (PA changed) sequentially at 34, 38, and 44 wk of age. Hens within the initially heterogeneous groups of 30, 50, and 70% marked birds remained unchanged and were used as controls. Groups of 40 individuals showed a reduction in BW gain and weekly hen-day-egg production after 30% PA changes, as compared with control counterparts. No differences were found in pens of 10 hens, and the groups of 20 showed intermediate results. A transient reduction in egg production was found after 50% PA changes. No further productive effects were observed after 70% changes. Our findings suggest that differences in hen appearance, which may occur due to variations in health status, injuries, and other natural causes, can be critical for production and welfare management practices depending both on the flock size and the birds' previous experience in exposure to group phenotypic heterogeneity.

  5. Study on the application of CANDLE burnup strategy to several nuclear reactors. JAERI's nuclear research promotion program, H13-002 (Contract research)

    International Nuclear Information System (INIS)

    The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. When this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the burnup of discharged fuel is about 40%. It means that the nuclear energy can be utilized for many hundreds years without new mining, enrichment and reprocessing, and the amount of spent fuel can be reduced considerably. However, in order to perform such a high fuel burnup some innovative technologies should be developed. Though development of innovative fuel will take a lot of time, intermediate re-cladding may be easy to be employed. Compared to fast reactors, application of CANDLE burnup to prismatic fuel high-temperature gas cooled reactors is very easy. In this report the application of CANDLE burnup to both these types of reactors are studied. (author)

  6. A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-06-15

    Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

  7. Study of Relationship between Anger, Self Efficiency, Coping Styles, Tendency to Narcotic Drug in a Group of Clients Dependent to Narcotic Drugs

    OpenAIRE

    Mohammad Oraki

    2011-01-01

    Introduction: Main purpose of the present research is the study of relationship between anger, self efficiency, coping styles, tendency to narcotic drug in a group of clients dependent to narcotic drugs. Methods: 180 men referred to Tehran City Aftab clinic who were diagnosed to be addicted on the basis of diagnostic criteria of DSMIV-TR and completed detoxification stage were selected by random sampling in accordance with an descriptive design .The clients were assessed with tests of self e...

  8. H-bond length dependence of the distortion in acid tetrahedron groups H nAsO 4 ( n=1-3)

    Science.gov (United States)

    Ichikawa, Mizuhiko

    1988-07-01

    The H-bond length O⋯O dependence of the distortions of the AsO 4 tetrahedra of acid arsenate groups H nAsO 4 ( n=1-3) have been studied by examining the correlations of the As sbnd O bond lengths, OAsO angles, O sbnd O lengths and Baur's distortion indices (DI) with O⋯O and comparing them with those of the corresponding H nPO 4 groups previously reported. The behavior of the O⋯O dependence of the AsO 4 distortion is analogous to that of PO 4 distortion as a whole, that is (a) all DI increase with increasing O⋯O, and correspondingly some of the individual or average As sbnd O lengths, OAsO angles and O sbnd O lengths correlate with O⋯O consistently; (b) the overall mean values of As sbnd O, OAsO and O sbnd O are reasonably constant in each H nAsO 4 type, most notably in the case of OAsO. The dependence of the PO 4 distortion on O⋯O can be described approximately by a model in which the P atom shifts away from the centroid of the regular tetrahedral framework, but the O⋯O dependence of the AsO 4 distortion is more complicated.

  9. Surface modification of bacterial cellulose nanofibers for property enhancement of optically transparent composites: dependence on acetyl-group DS.

    Science.gov (United States)

    Ifuku, Shinsuke; Nogi, Masaya; Abe, Kentaro; Handa, Keishin; Nakatsubo, Fumiaki; Yano, Hiroyuki

    2007-06-01

    Bacterial cellulose (BC) nanofibers were acetylated to enhance the properties of optically transparent composites of acrylic resin reinforced with the nanofibers. A series of BC nanofibers acetylated from degree-of-substitution (DS) 0 to 1.76 were obtained. X-ray diffraction profiles indicated that acetylation proceeded from the surface to the core of BC nanofibers, and scanning electron microscopy images showed that the volume of nanofibers increases by the bulky acetyl group. Since acetylation decreased the refractive index of cellulose, regular transmittance of composites comprised of 63% BC nanofiber was improved, and deterioration at 580 nm because of fiber reinforcement was suppressed to only 3.4%. Acetylation of nanofibers changed their surface properties and reduced the moisture content of the composite to about one-third that of untreated composite, although excessive acetylation increased hygroscopicity. Furthermore, acetylation was found to reduce the coefficient of thermal expansion of a BC sheet from 3 x 10(-6) to below 1 x 10(-6) 1/K.

  10. Coarse time-step integration method for burnup calculation of LWR lattice containing gadolinium-poisoned rods

    International Nuclear Information System (INIS)

    For the purpose of enhancing the efficiency of the burnup calculation of LWR lattice, two coarse time-step integration methods have been developed, both of which are to be used in combination with the ordinary Runge-Kutta-Gill method. It has been ensured through the numerical results of model problems simulating the depletion of 157Gd in a gadolinium-poisoned rod that the maximum time-step size allowed by the proposed methods is roughly 4 or 5 times larger than that achieved by the Predictor-Corrector method known as an effective coarse time-step method, and consequently that the proposed methods would reduce the computation time to a half or less when applied to an LWR lattice burnup calculation. The factor of reduction of computation time is still more significant if compared with other conventional methods such as the Runge-Kutta-Gill method etc. In addition, it has been demonstrated through their application to the LWR lattice physics code TGBLA that no appreciable error is observed over the range of time-step size up to 1GWd/t in the burnup calculation for a typical BWR lattice containing gadolinium-poisoned rods. Although the method development and verification presented here place emphasis on the cases of LWR lattice burnup, it is expected that the proposed methods would be applicable equally well to general problems dealing with the nuclide transmutation due to burnup. (author)

  11. Use of burnup credit in criticality evaluation for spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Chon, Je Keun; Kim, Jae Chun; Koh, Duck Joon; Kim Byung Tae [Nuclear Environment Technology Institute, Korea Electric Power Corporation, Taejon (Korea, Republic of)

    1999-07-01

    Boraflex is a polymer based material which is used as matrix to contain a neutron absorber material, boron carbide. In a typical spent fuel pool the irradiated Boraflex has been known as a significant source of silica. Since 1996, it was reported that elevated silica levels were measured in the Ulchin Unit 2 spent fuel pool water. Therefore, the Ulchin Unit 2 spent fuel storage racks were needed to be reanalyzed to allow storage of fuel assemblies with normal enrichments up to 5.0w/o U-235 in all storage cell locations using credit for burnup. The analysis does not take any credit for the presence of the spent fuel rack Boraflex neutron absorber panels. In region 2, the calculations were performed by assuming in an infinite radial array of storage cells. No credit is taken for axial or radial neutron leakage. The water in the spent fuel storage pool was assumed to be pure. In the evaluation of the Ulchin Unit 2 spent fuel storage pool, criticality analyses were performed with the CASMO-3 code. A reactivity uncertainty in the fuel depletion calculations was combined with other calculational uncertainty. The manufacturing tolerances were considered, as well. From the calculation, the acceptable burnup domain in region 2 of the spent fuel storage pool. where the curve identifies conditions of equal reactivity for various initial enrichments between 1.6w/o and 5.0w/o, was evaluated. In region 2, the maximum k{sub e}ff including all uncertainties, is 0.94648 for the enrichment-burnup combination from loading curve. (author)

  12. High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-02-01

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

  14. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J. C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  15. Use of burnup credit in criticality evaluation for spent fuel storage pool

    International Nuclear Information System (INIS)

    Boraflex is a polymer based material which is used as matrix to contain a neutron absorber material, boron carbide. In a typical spent fuel pool the irradiated Boraflex has been known as a significant source of silica. Since 1996, it was reported that elevated silica levels were measured in the Ulchin Unit 2 spent fuel pool water. Therefore, the Ulchin Unit 2 spent fuel storage racks were needed to be reanalyzed to allow storage of fuel assemblies with normal enrichments up to 5.0w/o U-235 in all storage cell locations using credit for burnup. The analysis does not take any credit for the presence of the spent fuel rack Boraflex neutron absorber panels. In region 2, the calculations were performed by assuming in an infinite radial array of storage cells. No credit is taken for axial or radial neutron leakage. The water in the spent fuel storage pool was assumed to be pure. In the evaluation of the Ulchin Unit 2 spent fuel storage pool, criticality analyses were performed with the CASMO-3 code. A reactivity uncertainty in the fuel depletion calculations was combined with other calculational uncertainty. The manufacturing tolerances were considered, as well. From the calculation, the acceptable burnup domain in region 2 of the spent fuel storage pool. where the curve identifies conditions of equal reactivity for various initial enrichments between 1.6w/o and 5.0w/o, was evaluated. In region 2, the maximum keff including all uncertainties, is 0.94648 for the enrichment-burnup combination from loading curve. (author)

  16. Latitude Dependence of the Variations of Sunspot Group Numbers (SGN) and Coronal Mass Ejections (CMEs) in Cycle 23

    Science.gov (United States)

    Kane, R. P.

    2008-06-01

    The 12-month running means of the conventional sunspot number Rz, the sunspot group numbers (SGN) and the frequency of occurrence of Coronal Mass Ejections (CMEs) were examined for cycle 23 (1996 - 2006). For the whole disc, the SGN and Rz plots were almost identical. Hence, SGN could be used as a proxy for Rz, for which latitude data are not available. SGN values were used for 5° latitude belts 0° - 5°, 5° - 10°, 10° - 15°, 15° - 20°, 20° - 25°, 25° - 30° and > 30°, separately in each hemisphere north and south. Roughly, from latitudes 25° - 30° N to 20° - 25° N, the peaks seem to have occurred later for lower latitudes, from latitudes 20° - 25° N to 15° - 20° N, the peaks are stagnant or occur slightly earlier, and then from latitudes 15° - 20° N to 0° - 5° N, the peaks seem to have occurred again later for lower latitudes. Thus, some latitudinal migration is suggested, clearly in the northern hemisphere, not very clearly in the southern hemisphere, first to the equator in 1998, stagnant or slightly poleward in 1999, and then to the equator again from 2000 onwards, the latter reminiscent of the Maunder butterfly diagrams. Similar plots for CME occurrence frequency also showed multiple peaks (two or three) in almost all latitude belts, but the peaks were almost simultaneous at all latitudes, indicating no latitudinal migration. For similar latitude belts, SGN and CME plots were dissimilar in almost all latitude belts except 10° - 20° S. The CME plots had in general more peaks than the SGN plots, and the peaks of SGN often did not match with those of CME. In the CME data, it was noticed that whereas the values declined from 2002 to 2003, there was no further decline during 2003 - 2006 as one would have expected to occur during the declining phase of sunspots, where 2007 is almost a year of sunspot minimum. An inquiry at GSFC-NASA revealed that the person who creates the preliminary list was changed in 2004 and the new person picks out

  17. Evaluation of a high burnup spent fuel regarding the regulations for a spent fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ik Sung; Yang, Young Sik; Bang, Je Geon; Kim, Dae Ho; Kim, Sun Ki; Song, Keun Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    All nuclear plants have storage pools for spent fuel. These pools are typically 40 or more feet deep. In many countries, the spent fuels are stored under water. The water serves 2 purposes: 1) It serves as a shield to reduce the radiation levels. 2) It cools the fuel assemblies that continue to produce heat (called decay heat). But Korean nuclear plant expects the storage capacity to reach its limit by the year 2016. So, the research for the spent fuel dry storage facilities is necessary. The purpose of this study was to overview the regulatory basis for spent fuel dry storage and to evaluate its applicability for high burnup spent fuel.

  18. Investigation of Irradiation Behavior of SiC-Coated Fuel Particle at Extended Burnup

    International Nuclear Information System (INIS)

    In current high-temperature gas-cooled reactors (HTGRs), Tri-isotropic (TRISO)-coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. The maximum burnup of the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR) is limited to 3.6%FIMA (% fission per initial metallic atom) to certify its integrity during the operation. In order to investigate fuel behavior under extended burnup condition, irradiation tests were performed. The irradiation was carried out as HRB-22 and 91F-1A capsule irradiation tests. The fuel for the irradiation tests was called extended burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the HTTR. In order to keep fuel integrity up to over 5%FIMA, the thickness of buffer and SiC layers of fuel particle were increased. The fuel compacts were irradiated in the HRB-22 and the 91F-1A capsules at the High Flux Isotope Reactor of Oak Ridge National Laboratory and at the Japan Materials Testing Reactor of the Japan Atomic Energy Research Institute, respectively. The comparison of measured and calculated release rate-to-birth rate ratios showed that there were additional failures in both irradiation tests. A pressure vessel failure model analysis showed that no tensile stresses acted on the SiC layers even at the end of irradiation and no pressure vessel failure occurred in the intact particles even in a particle with thin buffer layer with failed OPyC layer. The presumed failure mechanisms are additional through-coatings failure of as-fabricated SiC-failed particles or an excessive increase of internal pressure

  19. Calculation of the CB1 burnup credit benchmark reaction rates with MCNP4B

    International Nuclear Information System (INIS)

    The first calculational VVER-440 burnup credit benchmark CB1 in 1996. VTT Energy participated in the calculation of the CB1 benchmark with three different codes: CASMO-4, KENO-VI and MCNP4B. However, the reaction rates and the fission ν were calculated only with CASMO-4. Now, the neutron absorption and production reaction rates and the fission ν values have been calculated at VTT Energy with the MCNP4B Monte Carlo code using the ENDF60 neutron data library. (author)

  20. Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia

    OpenAIRE

    Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto

    2015-01-01

    A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the n...

  1. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  2. Analysis of the distribution and evolution of the ATP-dependent DNA ligases of bacteria delineates a distinct phylogenetic group 'Lig E'.

    Science.gov (United States)

    Williamson, Adele; Hjerde, Erik; Kahlke, Tim

    2016-01-01

    Prior to the discovery of a minimal ATP-dependent DNA ligase in Haemophilus influenzae, bacteria were thought to only possess a NAD-dependent ligase, which was involved in sealing of Okazaki fragments. We now know that a diverse range of bacterial species possess up to six of these accessory bacterial ATP-dependent DNA ligases (b-ADLs), which vary in size and enzymatic domain associations. Here we compare the domain structure of different types of b-ADLs and investigate their distribution among the bacterial domain to describe possible evolutionary trajectories that gave rise to the sequence and structural diversity of these enzymes. Previous biochemical and genetic analyses have delineated three main classes of these enzymes: Lig B, Lig C and Lig D, which appear to have descended from a common ancestor within the bacterial domain. In the present study, we delineate a fourth group of b-ADLs, Lig E, which possesses a number of unique features at the primary and tertiary structural levels. The biochemical characteristics, domain structure and inferred extracellular location sets this group apart from the other b-ADLs. The results presented here indicate that the Lig E type ligases were horizontally transferred into bacteria in a separate event from other b-ADLs possibly from a bacteriophage. PMID:26412580

  3. Dependence of the Sunspot-group Size on the Level of Solar Activity and its Influence on the Calibration of Solar Observers

    CERN Document Server

    Usoskin, I G; Chatzistergos, T

    2016-01-01

    The distribution of the sunspot group size (area) and its dependence on the level of solar activity is studied. It is shown that the fraction of small groups is not constant but decreases with the level of solar activity so that high solar activity is largely defined by big groups. We study the possible influence of solar activity on the ability of a realistic observer to see and report the daily number of sunspot groups. It is shown that the relation between the number of sunspot groups as seen by different observers with different observational acuity thresholds is strongly non-linear and cannot be approximated by the traditionally used linear scaling ($k-$factors). The observational acuity threshold [$A_{\\rm th}$] is considered to quantify the quality of each observer, instead of the traditional relative $k-$factor. A nonlinear $c-$factor based on $A_{\\rm th}$ is proposed, which can be used to correct each observer to the reference conditions. The method is tested on a pair of principal solar observers, Wo...

  4. Determination of burnup balance for nuclear reactor fuel on the basis of γ-spectrometric determination of fission products

    International Nuclear Information System (INIS)

    Results are given of experimental investigations in one of the versions of the method for determination of the balance of nuclear fuel burnup process by means of the γ-spectrometry of fission products. In the version being considered a balance of the burnup process was determined on the base of 106Ru, 134Cs.Activity was measured by means of a γ-spectrometer with Ge counter. Investigations were done on the natural uranium metal fuel from the heavy-water moderated reactor of the first Czechoslovakian nuclear power plant A1 in Yaslovske Bohunice. Possibility was checked of determination of the fuel burnup depth as well as of the isotope ratio and content of plutonium. Results were compared with the control data which had been obtained on the base of the mass-spectrometry of U, Pu and Nd. The reasors for deviations were estimated in the cases when they were greater tan error in the control data

  5. Development of destructive methods of burn-up determination and their application on WWER type nuclear fuels

    International Nuclear Information System (INIS)

    Results are described of a cooperation between the Central Institute of Nuclear Research Rossendorf and the Radium Institute 'V.G. Chlopin' Leningrad in the field of destructive burn-up determination. Laboratory methods of burn-up determination using the classical monitors 137Cs, 106Ru, 148Nd and isotopes of heavy metals (U, Pu) as well as the usefulness of 90Sr, stable isotopes of Ru and Mo as monitors are dealt with. The analysis of the fuel components uranium (spectrophotometry, potentiometric titration, mass-spectrometric isotope dilution) and plutonium (spectrophotometry, coulometric titration, mass- and alpha-spectrometric isotope dilution) is fully described. Possibilities of increasing the reproducibility (automatic adjusting of measurement conditions) and the sensibility (ion impuls counting) of mass-spectrometric measurements are proposed and applied to a precise determination of Am and Cm isotopic composition. The methods have been used for burn-up analysis of spent WWER (especially WWER-440) fuel. (author)

  6. Comparison of Computational Estimations of Reactivity Margin From Fission Products and Minor Actinides in PWR Burnup Credit

    International Nuclear Information System (INIS)

    This paper has presented the results of a computational benchmark and independent calculations to verify the benchmark calculations for the estimation of the additional reactivity margin available from fission products and minor actinides in a PWR burnup credit storage/transport environment. The calculations were based on a generic 32 PWR-assembly cask. The differences between the independent calculations and the benchmark lie within 1% for the uniform axial burnup distribution, which is acceptable. The Δk for KENO - MCNP results are generally lower than the other Δk values, due to the fact that HELIOS performed the depletion part of the calculation for both the KENO and MCNP results. The differences between the independent calculations and the benchmark for the non-uniform axial burnup distribution were within 1.1%

  7. Assesment of advanced step models for steady state Monte Carlo burnup calculations in application to prismatic HTGR

    Directory of Open Access Journals (Sweden)

    Kępisty Grzegorz

    2015-09-01

    Full Text Available In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5. The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.

  8. Preparation of uranium-plutonium carbide-based fuels simulating high burnup by carbothermic reduction and their properties

    International Nuclear Information System (INIS)

    Three types, hypostoichiometric, nearly stoichiometric and hyperstoichiometric, of uranium-plutonium carbide fuels simulating 10 at.% burnup were prepared by carbothermic reduction of oxide containing fission product elements. The carbides contained fission product phases such as the UMoC2 and the U2RuC2 type or the RECsub(1.5-2.0) phases (RE:rare earth). Composite theoretical densities of heterogenious carbides containing the UC, U2C3 type and fission product phases were calculated from the proportions and densities of these phases. By comparison of specific volume of the carbide between of 0 at.% and 10 at.% burnup, the solid fission product swelling rate of a carbide-based fuel was estimated to be 0.4-0.5 % per at.% burnup. (author)

  9. Anomalous role change of tertiary amino and ester groups as hydrogen acceptors in eudragit E based solid dispersion depending on the concentration of naproxen.

    Science.gov (United States)

    Ueda, Hiroshi; Wakabayashi, Shinobu; Kikuchi, Junko; Ida, Yasuo; Kadota, Kazunori; Tozuka, Yuichi

    2015-04-01

    Eudragit E (EGE) is a basic polymer incorporating tertiary amino and ester groups. The role of the functional groups of EGE in the formation of solid dispersion (SD) with Naproxen (NAP) was investigated. The glass transition temperature (Tg) of EGE decreased with the plasticizing effect of NAP up to 20% weight ratio. Addition of NAP at over 30% induced a rise in Tg, with the maximum value being reached at 60% NAP. Further addition of NAP led to a rapid drop of the Tg. A dramatic difference of physical stability between the SDs including 60 and 70% NAP was confirmed. The SD including 70% NAP rapidly crystallized at 40 °C with 75% relative humidity, while the amorphous state could be maintained over 6 months in the SD with 60% NAP. The infrared and (13)C solid state-NMR spectra of the SDs suggested a formation of ionic interaction between the carboxylic acid of NAP and the amino group of EGE. The SD with 20% NAP raised the (13)C spin-lattice relaxation (T1) of the amino group, but it decreased with over 30% NAP. The change in the (13)C-T1 disappeared with 70% NAP. The (13)C-T1 of the ester group rose depending on the amount of NAP. From these findings, we concluded that the role as hydrogen acceptor shifted from the amine to the ester group with an increase in amount of NAP. Furthermore, the amino group of EGE did not contribute to the interaction at over 70% NAP. These phenomena could be strongly correlated with Tg and stability.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2001-08-02

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

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

    International Nuclear Information System (INIS)

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

  12. Experimental techniques and results related to high burn-up investigations at the OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    Investigations of phenomena associated with extended or high burn-up comprise a considerable part of the fuel and materials research programme carried out at the OECD Halden Reactor Project, reflecting the interests and priorities of the nuclear industry in this area. For this purpose, sensors, re-instrumentation techniques, irradiation rigs and data evaluation methods have been developed and successfully applied which allow tests to be conducted at extended burn-up with high data quality. Re-instrumentation of base-irradiated fuel rods (50 MWd/kgU) with pressure transducers has been carried out with several rods, which were then subjected to high power, to study fission gas release. The results as well as those from another high burn-up test confirm the empirical fission gas release threshold developed at Halden, which predicts the onset of FGR to be about 1100 deg. C at this burn-up. The related interlinkage-resintering effects can be studied in special assemblies allowing sweep-out and analysis of fission products. Comprehensive investigations of fuel temperature data suggest a gradual degradation of UO2 thermal conductivity with burn-up. A special experiment is presently being irradiated to burn-up beyond 60 MWd/kgU, where the accuracy of temperature measurements is maintained by utilising expansion thermometers, insensitive to irradiation induced decalibration. A preliminary evaluation of the temperature increase points to a UO2 conductivity degradation of 35% at 50 MWd/kgUO2 and 600 deg. C. (author). 5 refs, 5 figs

  13. Evaluation and Selection of Boundary Isotopic Composition for Burnup Credit Criticality Safety Analysis of RBMK Spent Fuel Management

    International Nuclear Information System (INIS)

    The on-site wet-type spent fuel storage facility ISF-1 is currently used for interim storage of spent nuclear fuel removed from Chernobyl NPP power units. The results of ISF-1 preliminary criticality analyses demonstrated the need for using the burnup credit principle in nuclear safety analysis. This paper provides results from the selection and testing of computer codes for determining the isotopic composition of RBMK spent fuel. Assessment is carried out and conclusions are made on conservative approaches to fuel burnup credit in subsequent ISF-1 safety assessment. (author)

  14. Investigation of the CANLUB/sheath interface in CANDU fuel at extended burnup by XPS and SEM/WDX

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Behnke, R.; Duclos, A.M.; Gerwing, A.F. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Chan, P.K. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    1997-07-01

    A systematic investigation of the fuel-sheath interface in CANDU fuel as a function of extended burnup has been undertaken by XPS and SEM/WDX analysis. Adherent deposits of UO{sub 2} and fission products, including Cs, Ba, Rb, I, Te, Cd and possibly Ru, have been routinely identified on CANLUB coated and bare Zircaloy surfaces. Some trends in the distribution and chemistry of key fission products have begun to emerge. Several potential mechanisms for degradation of the CANLUB graphite layer at high burnup have been practically excluded. New evidence of carbon relocation within the fuel element and limited reaction with excess oxygen has also been obtained. (author)

  15. FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

    International Nuclear Information System (INIS)

    OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

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

    Energy Technology Data Exchange (ETDEWEB)

    Widiawati, Nina, E-mail: nina-widiawati28@yahoo.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10, Bandung (Indonesia)

    2015-09-30

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

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

    Science.gov (United States)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

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

  18. Burnup of Cadmium Decoupler Material in the Spallation Neutron Source Moderators

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, BD

    2001-08-21

    At the Spallation Neutron Source being constructed at Oak Ridge National Laboratory, power levels will be greater than at any other operating pulsed spallation neutron scattering facility. Some of the moderators at the facility will contain cadmium that will be used to tailor neutron time distributions by absorbing low-energy neutrons. Because of the higher operating power levels, indications are that there will be considerable burnup of this cadmium during the lifetime of the moderators. Cadmium burnup rates have been calculated for locations around the moderators. Assumed operating conditions for these calculations were a 2-mA beam of 1-GeV protons on the mercury target for 5,000 operating hours per year and a three-year lifetime for the moderators and inner-plug assembly. With the present proposed cadmium thickness in the moderator region (0.05 cm), Monte Carlo calculations indicate considerable depletion of the active cadmium isotope. In places, the calculations indicate complete depletion. An obvious solution to the problem would be to increase the cadmium thickness with a concomitant increase in heat load. Results from some cadmium heating calculations are also presented for a cadmium thickness of 0.05 cm.

  19. Propagation of uncertainty in system parameters of a LWR model by sampling MCNPX calculations - Burnup analysis

    International Nuclear Information System (INIS)

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

  20. Progress on burnup calculation methods coupling Monte Carlo and depletion codes

    Energy Technology Data Exchange (ETDEWEB)

    Leszczynski, Francisco [Comision Nacional de Energia Atomica, San Carlos de Bariloche, RN (Argentina). Centro Atomico Bariloche]. E-mail: lesinki@cab.cnea.gob.ar

    2005-07-01

    Several methods of burnup calculations coupling Monte Carlo and depletion codes that were investigated and applied for the author last years are described. here. Some benchmark results and future possibilities are analyzed also. The methods are: depletion calculations at cell level with WIMS or other cell codes, and use of the resulting concentrations of fission products, poisons and actinides on Monte Carlo calculation for fixed burnup distributions obtained from diffusion codes; same as the first but using a method o coupling Monte Carlo (MCNP) and a depletion code (ORIGEN) at a cell level for obtaining the concentrations of nuclides, to be used on full reactor calculation with Monte Carlo code; and full calculation of the system with Monte Carlo and depletion codes, on several steps. All these methods were used for different problems for research reactors and some comparisons with experimental results of regular lattices were performed. On this work, a resume of all these works is presented and discussion of advantages and problems found are included. Also, a brief description of the methods adopted and MCQ system for coupling MCNP and ORIGEN codes is included. (author)

  1. Chemical separation for the burnup determination of the U3Si/Al spent fuels

    International Nuclear Information System (INIS)

    The separation of U, Pu, and Nd for the burnup determination of the U3Si/Al spent fuel samples has been studied. The preliminary experiments were carried out with the simulated spent fuel solution. The solutions were prepared by adding of fission product elements to unirradiated U3Si/Al fuel samples. The fuel samples were dissolved in 6 M HNO3, 6 M HNO3 using mercury catalyst, or applying a mixture of HCl and HNO3 without any catalyst. All dissolved fuel solutions contained a small amount of a residue(silica). The trace silica reprecipitated from the fuel solutions taken for the separation was dissolved in HF and removed by subsequent evaporating to dryness. The separation of U and fission product elements from the various sample solutions was achieved by two sequential anion exchange resin separation procedures. The U, Pu and Nd can be purely isolated from the sample solutions with a large excess of Al by this chromatographic procedures. The dissolution and separation procedure used in this experiment were applied for burnup determination of real U3Si/Al spent fuels from HANARO reactor

  2. Recursive solutions for multi-group neutron kinetics diffusion equations in homogeneous three-dimensional rectangular domains with time dependent perturbations

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, Claudio Z. [Universidade Federal de Pelotas, Capao do Leao (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Bodmann, Bardo E.J.; Vilhena, Marco T. [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-graduacao em Engenharia Mecanica; Barros, Ricardo C. [Universidade do Estado do Rio de Janeiro, Nova Friburgo, RJ (Brazil). Inst. Politecnico

    2014-12-15

    In the present work we solve in analytical representation the three dimensional neutron kinetic diffusion problem in rectangular Cartesian geometry for homogeneous and bounded domains for any number of energy groups and precursor concentrations. The solution in analytical representation is constructed using a hierarchical procedure, i.e. the original problem is reduced to a problem previously solved by the authors making use of a combination of the spectral method and a recursive decomposition approach. Time dependent absorption cross sections of the thermal energy group are considered with step, ramp and Chebyshev polynomial variations. For these three cases, we present numerical results and discuss convergence properties and compare our results to those available in the literature.

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

    International Nuclear Information System (INIS)

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

  4. 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...

  5. Hydrocarbons depending on the chain length and head group adopt different conformations within a water-soluble nanocapsule: 1H NMR and molecular dynamics studies.

    Science.gov (United States)

    Choudhury, Rajib; Barman, Arghya; Prabhakar, Rajeev; Ramamurthy, V

    2013-01-10

    In this study we have examined the conformational preference of phenyl-substituted hydrocarbons (alkanes, alkenes, and alkynes) of different chain lengths included within a confined space provided by a molecular capsule made of two host cavitands known by the trivial name "octa acid" (OA). One- and two-dimensional (1)H NMR experiments and molecular dynamics (MD) simulations were employed to probe the location and conformation of hydrocarbons within the OA capsule. In general, small hydrocarbons adopted a linear conformation while longer ones preferred a folded conformation. In addition, the extent of folding and the location of the end groups (methyl and phenyl) were dependent on the group (H(2)C-CH(2), HC═CH, and C≡C) adjacent to the phenyl group. In addition, the rotational mobility of the hydrocarbons within the capsule varied; for example, while phenylated alkanes tumbled freely, phenylated alkenes and alkynes resisted such a motion at room temperature. Combined NMR and MD simulation studies have confirmed that molecules could adopt conformations within confined spaces different from that in solution, opening opportunities to modulate chemical behavior of guest molecules.

  6. Reevaluation of fuel enthalpy in NSRR test for high burnup fuels

    International Nuclear Information System (INIS)

    This paper describes the recent procedure of evaluation of the fuel enthalpy in the reactivity initiated accident (RIA) simulating tests performed at the nuclear safety research reactor (NSRR), and reports some important updates of the fuel enthalpies in the tests with high burnup PWR fuels. Previously, the fuel enthalpy had been evaluated by the procedure based on the short-life fission product measurement, i.e. a pellet slice was sampled from the test fuel rod after the NSRR test, a chemical separation process was applied to the solution of the pellet slice to separate barium, and the amount of Ba-140 was measured by gamma spectrometry on the separated barium. But a part of the results showed significant scattering even within the similar tests with similar fuels, which should have showed similar fuel enthalpies. The scattering appears to indicate the difficulty in treatment of the short-life nuclides after the completion of the NSRR test and unsuccessful measurement of the amount of fuel dissolved in the specimen preparation. Another difficulty of the procedure is that it is not repeatable for a specimen and so double check of an evaluation is not possible. Hence, an alternative procedure, which is based on the total amount of fissile materials evaluated by mass analysis, was developed and has been applied for the tests after 2003; the amount of fissile materials is input to a well-verified neutron transport calculation model for the NSRR reactor core to calculate a coupling factor of power densities between the test fuel rod and the NSRR driver fuel rods. This procedure does not require quickness and is repeatable, so it is applicable even many years later if the fuel sample is available. The recent procedure was thus applied to the tests before 2003, whose burnups are below 60 GWd/tU. It was shown that the fuel enthalpy had been significantly underestimated in the tests with high burnup PWR fuels: the test series HBO and TK. In this paper, the procedure

  7. Criticality safety evaluation for the direct disposal of used nuclear fuel. Preparation of data for burnup credit evaluation (Contract research)

    International Nuclear Information System (INIS)

    In the direct disposal of used nuclear fuel (UNF), criticality safety evaluation is one of the important issues since UNF contains some amount of fissile material. In the conventional criticality safety evaluation of UNF where the fresh fuel composition is conservatively assumed, neutron multiplication factor is becoming overestimated as the fuel enrichment increases. The recent development of higher-enrichment fuel has therefore enhanced the benefit of the application of burnup credit. When applying the burnup credit to the criticality safety analysis of the disposed fuel system, the safe-side estimation of the reactivity is required taking into account the factors which affect the neutron multiplication factor of the burnt fuel system such as the nuclide composition uncertainties. In this report, the effects of the several parameters on the reactivity of disposal canister model were evaluated for used PWR fuel. The parameters are relevant to the uncertainties of depletion calculation code, irradiation history, and axial and horizontal burnup distribution, which are known to be important effect in the criticality safety evaluation adopting burnup credit. The latest data or methodology was adopted in this evaluation, based on the various latest studies. The appropriate margin of neutron multiplication factor in the criticality safety evaluation for UNF can be determined by adopting the methodology described in the present study. (author)

  8. Kinetic parameters study based on burn-up for improving the performance of research reactor equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2014-01-01

    Full Text Available In this study kinetic parameters, effective delayed neutron fraction and prompt neutron generation time have been investigated at different burn-up stages for research reactor's equilibrium core utilizing low enriched uranium high density fuel (U3Si2-Al fuel with 4.8 g/cm3 of uranium. Results have been compared with reference operating core of Pakistan research Reactor-1. It was observed that by increasing fuel burn-up, effective delayed neutron fraction is decreased while prompt neutron generation time is increased. However, over all ratio beff/L is decreased with increasing burn-up. Prompt neutron generation time L in the understudy core is lower than reference operating core of reactor at all burn-up steps due to hard spectrum. It is observed that beff is larger in the understudy core than reference operating core of due to smaller size. Calculations were performed with the help of computer codes WIMSD/4 and CITATION.

  9. Study of morphology on the oxidation and the annealing of high burn-up UO{sub 2} spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Bang, Jae Geun; Yang, Yong Sik; Song, Keun Woo; Lee, Hyung Kwon; Kwon, Hyung Moon [KAERI, Daejon (Korea, Republic of)

    2005-12-15

    The morphology of the high burnup UO{sub 2} spent fuel, which was oxidized and annealed in a PIA (Post Irradiation Annealing) apparatus, has been observed. The high burnup fuel irradiated in Ulchin Unit 2, average rod burnup 57,000 MWd/tU, was transported to the KAERI's PIEF. The test specimen was used with about 200 mg of the spent UO{sub 2} fuel fragment of the local burnup 65,000 MWd/tU. This specimen was annealed at 1400 .deg. C for 4hrs after the oxidation for 3hrs to grain boundary using the PIA apparatus in a hot-cell. In order to oxidize the grain boundary, the oxidation temperature increased up to 500 .deg. C and held for 3hrs in the mixed gas (60 ml He and 100 ml STD-air) atmosphere. The amount of 85Kr during the whole test process was measured to know the fission gas release behavior using the online system of a beta counter and a gamma counter. The detailed micro-structure was observed by a SEM to confirm the change of the fuel morphology after this test. As the annealing temperature increased, the fission products were observed to move to the grain surface and grain boundary of the UO{sub 2} matrix. This specimen was re-structured through the reduction process, and the grain sizes were distributed from 5 to 10 {mu}m.

  10. Study on burnup credit evaluation method at JAERI towards securing criticality safety rationale for management of spent fuel

    International Nuclear Information System (INIS)

    A higher initial 235U enrichment is currently required in the nuclear fuel fabrication specification to realize higher fuel burnup. Traditionally, in the criticality safety design of spent fuel (SF) storage and transportation (S/T) casks or facilities, the fuel is usually assumed to be at its full initial enrichment (so called fresh fuel assumption) to provide a large safety allowance, which is sometimes excessively given, for example requiring unnecessarily large space between fuel assemblies. The burnup credit taken for criticality safety design is firstly implemented to the SF Storage Rack of Rokkasho Reprocessing Facility, which is completed and expected for operation soon. Except for that, no burnup credit has been taken in criticality safety design for SF S/T casks or intermediate storage facilities in Japan. Since in the near future it is considered inevitable to handle spent fuel massively, it is desired to implement the rational S/T design saving safety and economy by taking into account the fuel burnup in the criticality safety control. Computer codes and data which are vital to assess criticality safety in the design stage of nuclear fuel cycle facility have been developed and prepared to constitute a Japanese criticality safety handbook at JAERI

  11. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  12. Applicability of the MCNP-ACAB system to inventory prediction in high-burnup fuels: sensitivity/uncertainty estimates

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, N.; Cabellos, O. [Madrid Polytechnic Univ., Dept. of Nuclear Engineering (Spain); Cabellos, O.; Sanz, J. [Madrid Polytechnic Univ., 2 Instituto de Fusion Nuclear (Spain); Sanz, J. [Univ. Nacional Educacion a Distancia, Dept. of Power Engineering, Madrid (Spain)

    2005-07-01

    We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am{sup 242m} is Am{sup 241}(n,{gamma}-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am{sup 242m} concentration.The uncertainties in the inventory of fission products reach up to 30%.

  13. Fission gas release behavior in high burnup UO{sub 2} fuels with developed rim-structure

    Energy Technology Data Exchange (ETDEWEB)

    Une, Katsumi [Global Nuclear Fuel-Japan Co. Led., Oarai, Ibaraki (Japan); Kashibe, Shinji [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Hayashi, Kimio [Japan Atomic Energy Research Inst., Tokyo (Japan)

    2002-11-01

    The effect of rim structure formation and external restraint pressure on fission gas release at transient conditions has been examined by using an out-of-pile high pressure heating technique for high burnup UO{sub 2} fuels (60, 74 and 90 GWd/t), which had been irradiated in test reactors. The latter two fuels bore a developed rim structure. The maximum heating temperature was 1500 degC, and the external pressures were independently controlled in the range of 10-150 MPa. The present high burnup fuel data were compared with those of previously studied BWR fuels of 37 and 54 GWd/t with almost no rim structure. The fission gas release and bubble swelling due to the growth of grain boundary bubbles and coarsened rim bubbles were effectively suppressed by the strong restraint pressure of 150 MPa for all the fuels; however the fission gas release remarkably increased for the two high burnup fuels with the developed rim structure, even at the strong restraint conditions. From the stepwise de-pressurization tests at an isothermal condition of 1500degC, the critical external pressure, below which a large burst release due to the rapid growth and interlinkage of the bubbles abruptly begins, was increased from a 40-60 MPa level for the middle burnup fuels to a high level of 120-140 MPa for the rim-structured high burnup fuels. The high potential for transient fission gas release and bubble swelling in the rim-structured fuels was attributed to highly over-pressurized fission gases in the rim bubbles. (author)

  14. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    Science.gov (United States)

    Díez, C. J.; Cabellos, O.; Martínez, J. S.

    2015-01-01

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

  15. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  16. Flattening of burnup reactivity in long-life prismatic HTGR by particle type burnable poisons

    International Nuclear Information System (INIS)

    Highlights: ► The effect of particle-type burnable poisons in long-life prismatic HTGR was analyzed in detail. ► Different burnable poison particles can be combined to minimize excess reactivity during the core life. ► The use of burnable poison particles increases the passive safety features of prismatic HTGRs. - Abstract: For the flattening of burnup reactivity in long-life prismatic High-Temperature Gas-cooled Reactors (HTGRs), the effect of particle type on burnable poison properties is analyzed in detail using Monte Carlo calculations. Some examples of optimized specifications are shown. It is shown that combinations of particles with different materials, diameters, and concentrations make it possible to reduce excess reactivity to around or below 1 $ during the core life. The use of optimized burnable poison particles will help improve the passive safety features of long-life prismatic HTGR

  17. Determination of nuclear fuel burnup by non-destructive gamma spectroscopy

    International Nuclear Information System (INIS)

    The determination of nuclear fuel burnup by the non-destructive gamma spectroscopy method is studied. A MTR (Materials Testing Reactor) -type fuel element is used in the measurement. The fuel element was removed from the reactor core in 1958 and, because of the long decay time, show only one peak in is gamma spectrum at 661.6 Kev. Corresponding to 137Cs. Measurements are made at 330 points of the element using a Nal detector and the final result revealed that the quantity of 235U consumed was 3.3 +- 0,8 milligram in the entire element. The effect of the migration of 137Cs in the element is neglected in view of the fact that it occurs only when the temperature is above 10000C, which is not the case in IEAR-1. (Author)

  18. Computational fluid dynamics analysis for K24B cask design with burnup credit

    International Nuclear Information System (INIS)

    Korea Nuclear Engineering Service Corp. (KONES) has designed K24B cask for the storage and the transportation of 24 (CE-type 16x16) PWR assemblies. K24B cask is designed with considering burnup credit of spent fuel. In order to remove heat from the fuel assemblies effectively, the flow channels in the upper and the lower part of fuel assemblies are set up to promote the natural convection. Computational fluid dynamics analysis is carried out to estimate and assure the thermal integrity of K24B cask. Conduction and radiation heat transfer through the cask components and the natural convective heat transfer in the cask are simulated. As a result of the analysis, the maximum temperatures of the cask components are maintained below the operating temperature for the safety. Therefore, the design of K24B cask can satisfy the safety limit. (author)

  19. Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

    CERN Document Server

    Diez, Carlos Javier; Hoefer, Axel; Porsch, Dieter; Cabellos, Oscar

    2014-01-01

    Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact ...

  20. Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1980-08-01

    A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.

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

    Science.gov (United States)

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

    2016-08-01

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

  2. Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

    International Nuclear Information System (INIS)

    A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems

  3. Utilizing the burnup capability in MCNPX to perform depletion analysis of an MNSR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boafo, Emmanuel [Ghana atomic Energy Commission, Accra (Ghana)

    2013-07-01

    The burnup capability in the MCNPX code was utilized to perform fuel depletion analysis of the MNSR LEU core by estimating the amount of fissile material (U-235) consumed as well as the amount of plutonium formed after the reactor core expected life. The decay heat removal rate for the MNSR after reactor shutdown was also investigated due to its significance to reactor safety. The results show that 0.568 % of U-235 was burnt up after 200 days of reactor operation while the amount of plutonium formed was not significant. The study also found that the decay heat decreased exponentially after reactor shutdown confirming that the decay heat will be removed from the system by natural circulation after shut down and hence safety of the reactor is assured.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-01

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

  5. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    International Nuclear Information System (INIS)

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  6. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Jang, Jung Nam; Hwang, Yong Hwa; Kwon, In Chan; Min, Duck Kee; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

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

    Science.gov (United States)

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

    2012-12-01

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

  8. Investigation of the fundamental constants stability based on the reactor Oklo burn-up analysis

    CERN Document Server

    Onegin, M S

    2010-01-01

    The burn-up for SC56-1472 sample of the natural Oklo reactor zone 3 was calculated using the modern Monte Carlo codes. We reconstructed the neutron spectrum in the core by means of the isotope ratios: $^{147}$Sm/$^{148}$Sm and $^{176}$Lu/$^{175}$Lu. These ratios unambiguously determine the spectrum index and core temperature. The effective neutron absorption cross section of $^{149}$Sm calculated using this spectrum was compared with experimental one. The disagreement between these two values allows to limit a possible shift of the low laying resonance of $^{149}$Sm even more . Then, these limits were converted to the limits for the change of the fine structure constant $\\alpha$. We found that for the rate of $\\alpha$ change the inequality $|\\delta \\dot{\\alpha}/\\alpha| \\le 5\\cdot 10^{-18}$ is fulfilled, which is of the next higher order than our previous limit.

  9. Triton burnup study using scintillating fiber detector on JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Harano, Hideki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1997-09-01

    The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV {alpha} particles behave. The {alpha} particles` behavior can be simulated using the 1 MeV triton. To investigate the 1 MeV triton`s behavior, a new type of directional 14 MeV neutron detector, scintillating fiber (Sci-Fi) detector has been developed and installed on JT-60U in the cooperation with LANL as part of a US-Japan collaboration. The most remarkable feature of the Sci-Fi detector is that the plastic scintillating fibers are employed for the neutron sensor head. The Sci-Fi detector measures and extracts the DT neutrons from the fusion radiation field in high time resolution (10 ms) and wide dynamic range (3 decades). Triton burnup analysis code TBURN has been made in order to analyze the time evolution of DT neutron emission rate obtained by the Sci-Fi detector. The TBURN calculations reproduced the measurements fairly well, and the validity of the calculation model that the slowing down of the 1 MeV triton was classical was confirmed. The Sci-Fi detector`s directionality indicated the tendency that the DT neutron emission profile became more and more peaked with the time progress. In this study, in order to examine the effect of the toroidal field ripple on the triton burnup, R{sub p}-scan and n{sub e}-scan experiments have been performed. The R{sub p}-scan experiment indicates that the triton`s transport was increased as the ripple amplitude over the triton became larger. In the n{sub e}-scan experiment, the DT neutron emission showed the characteristic changes after the gas puffing injection. It was theoretically confirmed that the gas puffing was effective for the collisionality scan. (J.P.N.) 127 refs.

  10. ELESTRES 2.1 computer code for high burnup CANDU fuel performance analysis

    International Nuclear Information System (INIS)

    The ELESTRES (ELEment Simulation and sTRESses) computer code models the thermal, mechanical and micro structural behaviours of CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains in fuel element design analysis and assessments. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. ELESTRES 2.1 was developed for high burnup fuel application, based on an industry standard tool version of the code, through the implementation or modification to code models such as fission gas release, fuel pellet densification, flux depression (radial power distribution in the fuel pellet), fuel pellet thermal conductivity, fuel sheath creep, fuel sheath yield strength, fuel sheath oxidation, two dimensional heat transfer between the fuel pellet and the fuel sheath; and an automatic finite element meshing capability to handle various fuel pellet shapes. The ELESTRES 2.1 code design and development was planned, implemented, verified, validated, and documented in accordance with the AECL software quality assurance program, which meets the requirements of the Canadian Standards Association standard for software quality assurance CSA N286.7-99. This paper presents an overview of the ELESTRES 2.1 code with descriptions of the code's theoretical background, solution methodologies, application range, input data, and interface with other analytical tools. Code verification and validation results, which are also discussed in the paper, have confirmed that ELESTRES 2.1 is capable of modelling important fuel phenomena and the code can be used in the design assessment and the verification of high burnup fuels. (author)

  11. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations

    International Nuclear Information System (INIS)

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual keff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data

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

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

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

  13. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  14. On the use of SERPENT Monte Carlo code to generate few group diffusion constants

    Energy Technology Data Exchange (ETDEWEB)

    Piovezan, Pamela, E-mail: pamela.piovezan@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Carluccio, Thiago; Domingos, Douglas Borges; Rossi, Pedro Russo; Mura, Luiz Felipe, E-mail: fermium@cietec.org.b, E-mail: thiagoc@ipen.b [Fermium Tecnologia Nuclear, Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The accuracy of diffusion reactor codes strongly depends on the quality of the groups constants processing. For many years, the generation of such constants was based on 1-D infinity cell transport calculations. Some developments using collision probability or the method of characteristics allow, nowadays, 2-D assembly group constants calculations. However, these 1-D and 2-D codes how some limitations as , for example, on complex geometries and in the neighborhood of heavy absorbers. On the other hand, since Monte Carlos (MC) codes provide accurate neutro flux distributions, the possibility of using these solutions to provide group constants to full-core reactor diffusion simulators has been recently investigated, especially for the cases in which the geometry and reactor types are beyond the capability of the conventional deterministic lattice codes. The two greatest difficulties on the use of MC codes to group constant generation are the computational costs and the methodological incompatibility between analog MC particle transport simulation and deterministic transport methods based in several approximations. The SERPENT code is a 3-D continuous energy MC transport code with built-in burnup capability that was specially optimized to generate these group constants. In this work, we present the preliminary results of using the SERPENT MC code to generate 3-D two-group diffusion constants for a PWR like assembly. These constants were used in the CITATION diffusion code to investigate the effects of the MC group constants determination on the neutron multiplication factor diffusion estimate. (author)

  15. Group psycho-education in patients with bipolar disorder associated with a dependency of toxic substances in patients who are in abstinence

    Directory of Open Access Journals (Sweden)

    Patricia González Alegre

    2009-01-01

    Full Text Available The high comorbility that exists among psychiatric disorders and addictive is important. In the latest years it is produced an increase of the sensibility related to this problem. A great deal it is due to the demand of Mental Health Services and also due to drug dependency, as a consequence of the lack of an integral approach. Because of this fact and because of the mentioned demand, we though it should be pertinent developing a research project in order to check if the carrying the psycho-educative preventive group project out in patients with a diagnose of bipolar disorder with an abuse of drugs history and/or dependency of toxic substance in abstinence at the moment influents in a positive way in the course of the number of relapses in the toxic consumption during at least six months subsequent to the intervention. And at this way, these patients will purchase a greater consciousness of the important of healthy habits in the bipolar disorder and the recovery in the toxic substance abuse. The program will be developed in an experimental research where the patients will be randomly assigned in group control/ experimentally, the intervention will last twenty sessions, each session will be an hour and a half long and will be held weekly. In these sessions we will deal with topics related to the psychiatric disorder and the toxic consume. At the same time we will bank on the development of practical relaxation workshops on in some of the sessions with the object of providing a resource in view of stress situations.

  16. Dependency Analysis Guidance Nordic/German Working Group on Common Cause Failure analysis. Phase 2, Development of Harmonized Approach and Applications for Common Cause Failure Quantification

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Guenter; Johanson, Gunnar; Lindberg, Sandra; Vaurio, Jussi

    2009-03-15

    The Regulatory Code SSMFS 2008:1 of Swedish Radiation Safety Authority (SSM) includes requirements regarding the performance of probabilistic safety assessments (PSA), as well as PSA activities in general. Therefore, the follow-up of these activities is part of the inspection tasks of SSM. According to the SSMFS 2008:1, the safety analyses shall be based on a systematic identification and evaluation of such events, event sequences and other conditions which may lead to a radiological accident. The research report Nordic/German Working Group on Common cause Failure analysis. Phase 2 project report: Development of Harmonized Approach and Applications for Common Cause Failure Quantification has been developed under a contract with the Nordic PSA Group (NPSAG) and its German counterpart VGB, with the aim to create a common experience base for defence and analysis of dependent failures i.e. Common Cause Failures CCF. Phase 2 in this project if a deepened data analyses of CCF events and a demonstration on how the so called impact vectors can be constructed and on how CCF parameters are estimated. The word Guidance in the report title is used in order to indicate a common methodological guidance accepted by the NPSAG, based on current state of the art concerning the analysis of dependent failures and adapted to conditions relevant for Nordic sites. This will make it possible for the utilities to perform cost effective improvements and analyses. The report presents a common attempt by the authorities and the utilities to create a methodology and experience base for defence and analysis of dependent failures. The performed benchmark application has shown how important the interpretation of base data is to obtain robust CCF data and data analyses results. Good features were found in all benchmark approaches. The obtained experiences and approaches should now be used in harmonised procedures. A next step could be to develop and agree on event and formula driven impact vector

  17. Dependency Analysis Guidance Nordic/German Working Group on Common Cause Failure analysis. Phase 2, Development of Harmonized Approach and Applications for Common Cause Failure Quantification

    International Nuclear Information System (INIS)

    The Regulatory Code SSMFS 2008:1 of Swedish Radiation Safety Authority (SSM) includes requirements regarding the performance of probabilistic safety assessments (PSA), as well as PSA activities in general. Therefore, the follow-up of these activities is part of the inspection tasks of SSM. According to the SSMFS 2008:1, the safety analyses shall be based on a systematic identification and evaluation of such events, event sequences and other conditions which may lead to a radiological accident. The research report Nordic/German Working Group on Common cause Failure analysis. Phase 2 project report: Development of Harmonized Approach and Applications for Common Cause Failure Quantification has been developed under a contract with the Nordic PSA Group (NPSAG) and its German counterpart VGB, with the aim to create a common experience base for defence and analysis of dependent failures i.e. Common Cause Failures CCF. Phase 2 in this project if a deepened data analyses of CCF events and a demonstration on how the so called impact vectors can be constructed and on how CCF parameters are estimated. The word Guidance in the report title is used in order to indicate a common methodological guidance accepted by the NPSAG, based on current state of the art concerning the analysis of dependent failures and adapted to conditions relevant for Nordic sites. This will make it possible for the utilities to perform cost effective improvements and analyses. The report presents a common attempt by the authorities and the utilities to create a methodology and experience base for defence and analysis of dependent failures. The performed benchmark application has shown how important the interpretation of base data is to obtain robust CCF data and data analyses results. Good features were found in all benchmark approaches. The obtained experiences and approaches should now be used in harmonised procedures. A next step could be to develop and agree on event and formula driven impact vector

  18. Transport and Burnup Numerical Simulation on the Liquid Blanket Burnup of In-Zinerater%In-Zinerater液态包层输运燃耗数值模拟

    Institute of Scientific and Technical Information of China (English)

    师学明; 杨俊云; 刘成安

    2014-01-01

    Z-Pinch惯性约束聚变是未来一种有竞争力的能源候选方案。Z-Pinch驱动的聚变裂变混合堆可高效地嬗变反应堆乏燃料中分离出的超铀元素。对美国Sandia国家实验室提出的In-Zinerater混合堆概念进行了中子学分析和数值模拟。在三维输运燃耗耦合程序MCORGS中增加了处理在线添加燃料与去除裂变产物的功能,实现了对液态燃料燃耗过程的模拟。增加6Li丰度和燃料初装量保持寿期初反应性不变,可以减缓寿期内反应性下降趋势。逐步增加包层内超铀元素装量,可以控制整个寿期内反应性基本恒定。聚变功率取20 MW,通过反应性控制,5年内包层能量放大倍数在160∼180之间,氚增殖比在1.5∼1.7之间,优于In-Zinerater基准设计方案。%Z-Pinch Inertial confinement fusion is a competitive candidate for future energy solution. A fusion-fission hybrid driven by Z-Pinch can be used to transmute transuranic elements from spent fuels of reactors efficiently. Analysis and numerical simulation of blanket neutronics of In-Zinerater, which is a fusion-fission hybrid concept design in Sandia National Laboratories, is given in this paper. Modification to the three dimension transport and burnup code MCORGS are done, so as to simulate continuous feeding and continuous chemical processing of the liquid fuel. Different combination of initial enrichment of 6Li and fuels loading in the blanket are selected to keep the same reactivity at begin of core. By this way, the decreasing trend of reactivity at life of the core can be lowered. The reactivity can be maintained constant by increasing the fuel loading in the core gradually as the burnup deepens. Given a 20 MW fusion power, by reactivity control, the blanket energy multiplication is around 160∼180 and tritium breed ratio 1.5∼1.7 in 5 years, which is a better result than Sandia’s original design.

  19. Anisotropies in the Neutrino Fluxes and Heating Profiles in Two-dimensional, Time-dependent, Multi-group Radiation Hydrodynamics Simulations of Rotating Core-Collapse Supernovae

    CERN Document Server

    Walder, R; Ott, C D; Livne, E; Jarrah, M

    2004-01-01

    Using the 2D multi-group, flux-limited diffusion version of the code VULCAN/2D, that also incorporates rotation, we have calculated the collapse, bounce, shock formation, and early post-bounce evolutionary phases of a core-collapse supernova for a variety of initial rotation rates. This is the first series of such multi-group calculations undertaken in supernova theory with fully multi-D tools. We find that though rotation generates pole-to-equator angular anisotropies in the neutrino radiation fields, the magnitude of the asymmetries is not as large as previously estimated. Moreover, we find that the radiation field is always more spherically symmetric than the matter distribution, with its plumes and convective eddies. We present the dependence of the angular anisotropy of the neutrino fields on neutrino species, neutrino energy, and initial rotation rate. Only for our most rapidly rotating model do we start to see qualitatively different hydrodynamics, but for the lower rates consistent with the pre-collap...

  20. Burnup Credit of French PWR-MOx fuels: methodology and associated conservatisms with the JEFF-3.1.1 evaluation

    International Nuclear Information System (INIS)

    Considering spent fuel management (storage, transport and reprocessing), the approach using 'fresh fuel assumption' in criticality-safety studies results in a significant conservatism in the calculated value of the system reactivity. The concept of Burnup Credit (BUC) consists in considering the reduction of the spent fuel reactivity due to its burnup. A careful BUC methodology, developed by CEA in association with AREVA-NC was recently validated and written up for PWR-UOx fuels. However, 22 of 58 French reactors use MOx fuel, so more and more irradiated MOx fuels have to be stored and transported. As a result, why industrial partners are interested in this concept is because taking into account this BUC concept would enable for example a load increase in several fuel cycle devices. Recent publications and discussions within the French BUC Working Group highlight the current interest of the BUC concept in PWR-MOx spent fuel industrial applications. In this case of PWR-MOx fuel, studies show in particular that the 15 FPs selected thanks to their properties (absorbing, stable, non-gaseous) are responsible for more than a half of the total reactivity credit and 80% of the FPs credit. That is why, in order to get a conservative and physically realistic value of the application keff and meet the Upper Safety Limit constraint, calculation biases on these 15 FPs inventory and individual reactivity worth should be considered in a criticality-safety approach. In this context, thanks to an exhaustive literature study, PWR-MOx fuels particularities have been identified and by following a rigorous approach, a validated and physically representative BUC methodology, adapted to this type of fuel has been proposed, allowing to take fission products into account and to determine the biases related to considered isotopes inventory and to reactivity worth. This approach consists of the following studies: - isotopic correction factors determination to guarantee the criticality