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Sample records for burnup profiles npp

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

    International Nuclear Information System (INIS)

    Neuber, Jens-Christian; Lamprecht, Thomas

    1999-01-01

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

  2. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

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

  3. Triton burnup in JET - profile effects

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  4. Triton burnup in JET - profile effects

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  6. Results of calculation of WWER-440 fuel rods (Kol`skaya-3 NPP) at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Scheglov, A; Proselkov, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Panin, M; Pitkin, Yu [Kol` skaya NPP, (Russian Federation); Tzibulya, V [AO Mashinostroitelnij Zavod Electrostal (Russian Federation)

    1994-12-31

    Thermal-physical characteristics of fuel rods of two fuel assemblies which were operated within 5 - 8 and 5 - 9 core fuel loadings of the Unit 3 of the Kol`skaya NPP are calculated. They have achieved deep burnup during 4-year (> 46 Mwd/kg U) and 5-year (> 48 Mwd/kg U) fuel cycle. Fuel assemblies have been unloaded off the reactor and subjected to a post-irradiation testing. PIN-mod2 code originally designed for modelling of WWER fuel rod behaviour in a quasi-steady-state operation is used. The average fuel rod in the fuel assembly and the fuel rod with maximum burnup are selected. The preliminary comparison of the calculation results with those of the post-irradiation examination shows a satisfactory agreement. On the basis of the results obtained in the post-irradiation experiments an improvement of the model for calculation of fission gas release and creep of the cladding is planned. The results of the analysis performed indicate that the fuel rod completely preserves its working ability; fuel temperature does not exceed 1300{sup o} C; fission gas release does not exceed 4%; maximum gas pressure inside the cladding at the end of campaign does not exceed 2 MPa. 2 tabs., 11 figs., 5 refs.

  7. Impact of axial burnup profile on criticality safety of ANPP spent fuel cask

    International Nuclear Information System (INIS)

    Bznuni, S.

    2006-01-01

    Criticality safety assessment for WWER-440 NUHOMS cask with spent nuclear fuel from Armenian NPP has been performed. The cask was designed in such way that the neutron multiplication factor k eff must be below 0,95 for all operational modes and accident conditions. Usually for criticality analysis, fresh fuel approach with the highest enrichment is taken as conservative assumption as it was done for ANPP. NRSC ANRA in order to improve future fuel storage efficiency initiated research with taking into account burn up credit in the criticality safety assessment. Axial burn up profile (end effect) has essential impact on criticality safety justification analysis. However this phenomenon was not taken into account in the Safety Analysis Report of NUHOMS spent fuel storage constructed on the site of ANPP. Although ANRA does not yet accept burn up credit approach for ANPP spent fuel storage, assessment of impact of axial burnup profile on criticality of spent fuel assemblies has important value for future activities of ANRA. This paper presents results of criticality calculations of spent fuel assemblies with axial burn up profile. Horizontal burn up profile isn't taken account since influence of the horizontal variation of the burn up is much less than the axial variation. The actinides and actinides + fission products approach are discussed. The calculations were carried out with STARBUCS module of SCALE 5.0 code package developed at Oak Ridge National laboratory. SCALE5.0 sequence CSAS26 (KENO-VI) was used for evaluation the k eff for 3-D problems. Obtained results showed that criticality of ANPP spent fuel cask is very sensitive to the end effect

  8. Overview of the SEU project for extended burnup at the Atucha-I NPP. Four years of operating experience

    International Nuclear Information System (INIS)

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

    2002-01-01

    Atucha I is a 357 MWe nuclear station moderated and cooled with heavy water, of German design located in Argentina. Fuelling is on-power and the plant was originally fuelled with natural uranium. To reduce fuel costs a program was initiated in August 1993 to introduce gradually slightly enriched uranium (SEU) fuel (0.85 w% U-235) with an associated burnup increase from 5900 MWd/tU to 11300 MWd/tU. The introduction of SEU fuel started in January 1995 and the program was divided in three Phases with an upper limit of SEU FA in the core: 12, 60 and 252 (full core) and licensing documentation was prepared for each Phase. This paper describes the most important aspects of the operating and project experience, and some factors limiting the burnup extension from an operation point of view. After four years of the program and with 181 SEU FA (71%) of the core, the operating experience has been good and without unfavourable effects due to the use of SEU fuel with the only exception of a small increase of the time to reach full power in plant startups or power cycling. In particular, the new criteria to prevent PCI failures in power ramps for higher burnup SEU fuel in refueling operations, plant startups or power cycling has been effective. The average discharge burnup of the SEU fuel taken out of the reactor in 1998 was 11263 MWd/tU. The average discharge burnup of the natural fuel in the same year was 6640 MWd/tU, with an increase of about 12% of the original value for a natural fuel core. The average number of fresh fuel assemblies per full power day was being reduced from 1.31 to 0.92 in 1998 and 0.83 in 1999. The fuel costs dropped gradually during the program from 9.38 (with natural uranium fuel) to 6.57 $/MWh in the first four months of 1999 (taking as reference the NU and SEU FA costs for 1999). Because of this the SEU program has been an important contribution to the reduction of Atucha I operating costs and to the competitiveness of nuclear power generation against

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  10. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    Energy Technology Data Exchange (ETDEWEB)

    Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

    2016-10-15

    Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

  11. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    International Nuclear Information System (INIS)

    Barkauskas, V.; Plukiene, R.; Plukis, A.

    2016-01-01

    Highlights: • RBMK-1500 fuel burn-up impact on k_e_f_f in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k_e_f_f in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k_e_f_f) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

  12. The profile of quantitative risk indicators in Krsko NPP

    International Nuclear Information System (INIS)

    Vrbanic, I.; Basic, I.; Bilic-Zabric, T.; Spiler, J.

    2004-01-01

    During the past decade strong initiative was observed which was aimed at incorporating information on risk into various aspects of operation of nuclear power plants. The initiative was observable in activities carried out by regulators as well as utilities and industry. It resulted in establishing the process, or procedure, which is often referred to as integrated decision making or risk informed decision making. In this process, engineering analyses and evaluations that are usually termed traditional and that rely on considerations of safety margins and defense in depth are supplemented by quantitative indicators of risk. Throughout the process, the plant risk was most commonly expressed in terms of likelihood of events involving damage to the reactor core and events with radiological releases to the environment. These became two commonly used quantitative indicators or metrics of plant risk (or, reciprocally, plant safety). They were evaluated for their magnitude (e.g. the expected number of events per specified time interval), as well as their profile (e.g. the types of contributing events). The information for quantitative risk indicators (to be used in risk informing process) is obtained from plant's probabilistic safety analyses or analyses of hazards. It is dependable on issues such as availability of input data or quality of model or analysis. Nuclear power plant Krsko has recently performed Periodic Safety Review, which was a good opportunity to evaluate and integrate the plant specific information on quantitative plant risk indicators and their profile. The paper discusses some aspects of quantitative plant risk profile and its perception.(author)

  13. Comparative calculations and operation-to-PIE data juxtaposition of the Zaporozhye NPP, WWER-1000 FA-E0325 fuel rods after 4 years of operation up to ∼49 MWd/kgU burnup

    International Nuclear Information System (INIS)

    Passage, G.; Stefanova, S.; Scheglov, A.; Proselkov, V.

    2006-01-01

    Operational and PIE data for the Zaporozhe NPP, FA-E0325, WWER-1000 fuel rods were provided in the OECD NEA IFPE Database and were used to perform comparative calculations among several fuel performance codes. The fuel rods had been irradiated for 4 years of operation up to ∼49 MWd/kg U burnup. The fuel rod operation histories are developed for the PINw99, TRANSURANUS (V1M1J03) and TOPRA-2 codes. The initial state fuel rod parameters are analysed and calculations are carried out. The PIE data enable the comparison of experimental measurement with code-calculated values for cladding elongation (49 rods), FGR and gas pressure (35 rods). Cladding diameter creep-down and gap closure results are juxtaposed as well. The capability of the applied codes correctly to predict the WWER fuel rod performance is shown. The WWER-1000 fuel rod data include initial geometrical and design parameters of the fuel rods, as well as description of the operation regime, NPP unit loading history and PIE results at normal conditions. The data are sufficient for modelling all 312 fuel rod and for comparison of calculations with experimental results for a limited number of fuel rods. The comparison between the calculated and measured results discussed in this paper shows that the codes PINw99, TRANSURANUS and TOPRA-2, are capable of adequate predicting the thermophysical and the mechanical performance of the WWER-1000 fuel rods. The PINw99 code predicts conservative BOL FGR values and conservative gas pressure values in the region of burnups higher than 30 MWd/kg U, which can be explained by the underprediction of the cladding gas inner volume and cladding elongation. The improved version PIN2K (not applied in the present study) predicts much better FGR and gas pressure, though, it is still under development in the high burnup FGR modelling part. In the TRANSURANUS code, there are also areas, where refinements are clearly indicated. They are subjects of the ongoing research projects and

  14. An Algorithm For Climate-Quality Atmospheric Profiling Continuity From EOS Aqua To Suomi-NPP

    Science.gov (United States)

    Moncet, J. L.

    2015-12-01

    We will present results from an algorithm that is being developed to produce climate-quality atmospheric profiling earth system data records (ESDRs) for application to hyperspectral sounding instrument data from Suomi-NPP, EOS Aqua, and other spacecraft. The current focus is on data from the S-NPP Cross-track Infrared Sounder (CrIS) and Advanced Technology Microwave Sounder (ATMS) instruments as well as the Atmospheric InfraRed Sounder (AIRS) on EOS Aqua. The algorithm development at Atmospheric and Environmental Research (AER) has common heritage with the optimal estimation (OE) algorithm operationally processing S-NPP data in the Interface Data Processing Segment (IDPS), but the ESDR algorithm has a flexible, modular software structure to support experimentation and collaboration and has several features adapted to the climate orientation of ESDRs. Data record continuity benefits from the fact that the same algorithm can be applied to different sensors, simply by providing suitable configuration and data files. The radiative transfer component uses an enhanced version of optimal spectral sampling (OSS) with updated spectroscopy, treatment of emission that is not in local thermodynamic equilibrium (non-LTE), efficiency gains with "global" optimal sampling over all channels, and support for channel selection. The algorithm is designed for adaptive treatment of clouds, with capability to apply "cloud clearing" or simultaneous cloud parameter retrieval, depending on conditions. We will present retrieval results demonstrating the impact of a new capability to perform the retrievals on sigma or hybrid vertical grid (as opposed to a fixed pressure grid), which particularly affects profile accuracy over land with variable terrain height and with sharp vertical structure near the surface. In addition, we will show impacts of alternative treatments of regularization of the inversion. While OE algorithms typically implement regularization by using background estimates from

  15. On the organization of commercial production of profiled polyethylene sheets used for water proofing under NPP construction

    International Nuclear Information System (INIS)

    Likhachev, V.D.; Korenev, K.I.; Chukvaidze, K.I.; Dzhurinskij, M.B.

    1986-01-01

    The problem on the organization of commercial production of profiled polyethylene sheets with special longitudinal ribs which are anchorized in concrete is considered. The use of new water proofing material applied under NPP construction allows one to enhance the commercialization of water proofing works which ensured cost reduction of these works

  16. Identifying measures to balance the risk profile of the Tihange 2 NPP

    International Nuclear Information System (INIS)

    D'Eer, A.M.; Monniez, J.J.

    2001-01-01

    In Belgium, each Nuclear Power Plant is subject to a periodic safety reassessment. In this context, it was found to be desirable to perform a Probabilistic Safety Assessment (PSA) in support of the ten yearly back-fitting process. The Tihange 2 NPP is a 3-loop PWR having a thermal capacity of 2905 MW. Analysis of the plant's risk profile shows that implementing feasible measures for improvement of the shutdown risk, would be beneficial. This is because a configuration leading to significant risk, namely cold pressurization when the residual heat removal system is lost during reduced primary inventory, thus can be avoided. As a result the risk between reactor shutdown and power operation will be balanced. The presentation describes the lessons learnt regarding the Tihange 2 shutdown PSA model and the expected benefits following implementation of one of the proposed measures. (author)

  17. High burnup MOX fuel assembly

    International Nuclear Information System (INIS)

    Blanpain, P.; Brunel, L.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Nakamura, Jinichi; Suzuki, Motoe

    2001-01-01

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

  19. Enhancement and evaluation of an algorithm for atmospheric profiling continuity from Aqua to Suomi-NPP

    Science.gov (United States)

    Lipton, A.; Moncet, J. L.; Payne, V.; Lynch, R.; Polonsky, I. N.

    2017-12-01

    We will present recent results from an algorithm for producing climate-quality atmospheric profiling earth system data records (ESDRs) for application to data from hyperspectral sounding instruments, including the Atmospheric InfraRed Sounder (AIRS) on EOS Aqua and the Cross-track Infrared Sounder (CrIS) on Suomi-NPP, along with their companion microwave sounders, AMSU and ATMS, respectively. The ESDR algorithm uses an optimal estimation approach and the implementation has a flexible, modular software structure to support experimentation and collaboration. Data record continuity benefits from the fact that the same algorithm can be applied to different sensors, simply by providing suitable configuration and data files. Developments to be presented include the impact of a radiance-based pre-classification method for the atmospheric background. In addition to improving retrieval performance, pre-classification has the potential to reduce the sensitivity of the retrievals to the climatological data from which the background estimate and its error covariance are derived. We will also discuss evaluation of a method for mitigating the effect of clouds on the radiances, and enhancements of the radiative transfer forward model.

  20. High burnup issues and modelling strategies

    International Nuclear Information System (INIS)

    Dutta, B.K.

    2005-01-01

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

  1. Nuclear fuel burn-up economy

    International Nuclear Information System (INIS)

    Matausek, M.

    1984-01-01

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

  2. Choosing the optimum burnup

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  3. Economic aspects of Dukovany NPP fuel cycle

    International Nuclear Information System (INIS)

    Vesely, P.; Borovicka, M.

    2001-01-01

    The paper discusses some aspects of high burnup program implementation at Dukovany NPP and its influence on the fuel cycle costs. Dukovany internal fuel cycle is originally designed as a three years cycle of the Out-In-In fuel reloading patterns. These reloads are not only uneconomical but they additionally increased the radiation load of the reactor pressure vessel due to high neutron leakage typical for Out-In-In loading pattern. To avoid the high neutron leakage from the core a transition to 4-year fuel cycle is started in 1987. The neutron leakage from the core is sequentially decreased by insertion of older fuel assemblies at the core periphery. Other developments in fuel cycle are: 1) increasing of enrichment in control assemblies (3.6% of U-235); 2) improvement in fuel assembly design (reduce the assembly shroud thickness from 2.1 to 1.6 mm); 3) introduction of Zr spacer grid instead of stainless steel; 4) introduction of new type of assembly with profiled enrichment with average value of 3.82%. Due to increased reactivity of the new assemblies the transition to the partial 5-year fuel cycle is required. Typical fuel loading pattern for 3, 3.5, 4 and 5-year cycles are shown in the presented paper. An evaluation of fuel cost is also discussed by using comparative analysis of different fuel cycle options. The analysis shows that introduction of the high burnup program has decrease relative fuel cycle costs

  4. Lattice cell burnup calculation

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1977-01-01

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

  5. Physical models for high burnup fuel

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  6. High Burnup Effects Program

    International Nuclear Information System (INIS)

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

    1990-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-07-01

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

  8. Appropriate burnup measurements for transportation burnup credit

    International Nuclear Information System (INIS)

    Lancaster, D.; Fuentes, E.

    1997-01-01

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

  9. The relevance of axial burn-up profiles for the criticality safety analysis of spent nuclear fuel in a final repository

    International Nuclear Information System (INIS)

    Kilger, R.; Gmal, B.; Moser, E.F.

    2008-01-01

    Due to inhomogeneous neutron flux and moderator density distributions in the reactor core, the burn-up of a nuclear fuel assembly is not homogeneous but shows an axial distribution, typically with lower partial burn-up and thus higher remaining reactivity at the fuel ends in particular at the assembly top end. Beyond a burn-up of about 15 to 20 GWd/tHM, the multiplication factor K of the whole assembly is dominated by this lower-burnt end regions, and is usually higher than for assuming a homogeneous uniform distribution of the averaged burn-up. This behaviour commonly referred to as positive ''end effect'' is well known in burn-up credit considerations for transportation and storage casks and is being investigated also in the context of criticality analyses for final disposition of spent nuclear fuel. Sign and value of the end effect depend on several parameters. Based on a generic model one may not conclude that criticality in a final repository is a likely or expected event, but nevertheless it draws the attention to the fact that criticality is not excluded per se but has to be considered in the analysis and probably has to be encountered by certain appropriate measures, maybe e.g. by limitation of the amount of fissile material inside one single cask, or a rigorous prove for prevention of water ingress. The authors also conclude that the higher partial reactivity of the fuel ends has to be accounted for carefully in more realistic analyses of post-closure scenarios with respect to criticality safety.

  10. Burnup credit in Spain

    International Nuclear Information System (INIS)

    Conde, J.M.; Recio, M.

    2001-01-01

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

  11. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Kang, C.; Lancaster, D.

    1997-11-01

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

  13. Triton burnup in JET

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    BSC

    2004-01-01

    , initial 235 U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed

  15. COGEMA/TRANSNUCLEAIRE's experience with burnup credit

    International Nuclear Information System (INIS)

    Chanzy, Y.; Guillou, E.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  17. Belene NPP

    International Nuclear Information System (INIS)

    Tsvetanov, P.

    1990-01-01

    The book presents the main results of the studies of the Bulgarian Academy of Sciences (BAS) on the construction of a new nuclear power plant at Belene on the Danube river. The programme of the studies comprises five areas: the socio-economic and energy development and the necessity of the commissioning; a technical project and design level of the equipment (safety, radioactivity control, waste disposal and economic efficiency of the power plant); the seismic properties of the construction site; the corresponding risk and design features of the plant; the ecological impacts of the NPP and public opinion. The studies in the different areas have been carried out by independent teams, fully responsible for the formulated topical conclusions. The general opinion of the BAS voiced in the book is that the construction of Belene NPP is not sufficiently substantiated and is considered unacceptable. 94 refs., 53 fig., 56 tabs. (R.Ts.)

  18. ABB high burnup fuel

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  19. Modelling the high burnup UO2 structure in LWR fuel

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. EVOLUT - a computer program for fast burnup evaluation

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  1. Increased burnup of fuel elements

    International Nuclear Information System (INIS)

    Ahlf, J.

    1983-01-01

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

  2. Research on burnup physics

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-07-01

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

  3. Heat and radiation analysis of NPP Krsko irradiated fuel

    International Nuclear Information System (INIS)

    Lalovic, M.

    1986-01-01

    Radioactive and heat potential for irradiated fuel in the region 2 with burnup of 13400 MWd/tHM, and in the region 4A with burnup of 9360 MWd/tHM for NPP KRSKO, was calculated. Computer code KORIGEN (Karlsruhe Oak Ridge Isotope Generation and Depletion Code) was used. The aspects of radiation (mainly gamma and neutrons) and of heat production was considered with respect to their impact on fuel handing and waste management. Isotopic concentrations for irradiated fuel was calculated and compared with Westinghouse data. (author)

  4. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    Huffer, J.

    2004-01-01

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

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

    International Nuclear Information System (INIS)

    Miro, R.; Verdu, G.; Munoz-Cobo, J. L.

    1998-01-01

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

  6. The evolution of the fuel cycle in the Dukovany NPP

    Energy Technology Data Exchange (ETDEWEB)

    Bajgl, J [Jaderna Elektrarna, Dukovany (Czech Republic)

    1994-12-31

    The ten-year operational experience of four WWER-440 units in Dukovany NPP with a total number of 35 cycles is outlined. The strategy of fuel reloading has been changed from out-in schemes to low-leakage patterns. The linear pin power limitation will be introduced. The main physical limits and conditions for the NPP operation are listed. The main goal is to go to a full 4-year fuel cycle in which the burnup will be about 40 Mwd/kg U. 6 tabs.

  7. Decommissioning of NPP A-1

    International Nuclear Information System (INIS)

    Anon

    2009-01-01

    In this presentation the Operation history of A1 NPP, Project 'Decommissioning of A1 NPP' - I stage, Project 'Decommissioning of A1 NPP ' - II stage and Next stages of Project 'Decommissioning of A1 NPP ' are discussed.

  8. Improvements for Monte Carlo burnup calculation

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  9. Special aspects of implementing advanced fuel cycles at Kalinin NPP

    International Nuclear Information System (INIS)

    Tsvetkov, A.

    2015-01-01

    The presentation showed the experience of different TVSA modifications usage at Kalinin NPP. The strategy of 18 month fuel cycles implementation at uprated power (104%) was also presented. The transition and equilibrium fuel loadings features were discussed. The implementation of burn-up measurement installation MKS-01 was presented, in order to solve the spent nuclear fuel handling and transportation issues due to the increased fuel enrichment and heavy metal mass

  10. The radial distribution of plutonium in high burnup UO2 fuels

    International Nuclear Information System (INIS)

    Lassmann, K.; O'Carroll, C.; Laar, J. van de; Walker, C.T.

    1994-01-01

    A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21 000 and 64 000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions. (orig.)

  11. Mochovce NPP simulator

    International Nuclear Information System (INIS)

    Ziakova, M.

    1998-01-01

    Mochovce NPP simulator basic features and detailed description of its characteristics are presented with its performance, certification and application for training of NPP operators as well as the training scenario

  12. Effect of burnup history by moderator density on neutron-physical characteristics of WWER-1000 core

    International Nuclear Information System (INIS)

    Ovdiienko, I.; Kuchin, A.; Khalimonchuk, V.; Ieremenko, M.

    2011-01-01

    Results of assessment of burnup history effect by moderator density on neutron physical characteristics of WWER-1000 core are presented on example of stationary fuel loading with Russian design fuel assembly TWSA and AER benchmark for Khmelnitsky NPP that was proposed by TUV and SSTC NRC at nineteenth symposium. Assessment was performed by DYN3D code and cross section library sets generated by HELIOS code. Burnup history was taken into account by preparing of numerous cross section sets with different isotopic composition each of which was obtained by burning under different moderator density. For analysis of history effect 20 cross section sets were prepared for each fuel assembly corresponded to each of 20 axial layers of reactor core model for DYN3D code. Four fuel cycles were modeled both for stationary fuel loading with TWSA and AER benchmark for Khmelnitsky NPP to obtain steady value of error due to neglect of burnup history effect. Main attention of study was paid to effect of burnup history by moderator density to axial power distribution. Results of study for AER benchmark were compared with experimental values of axial power distribution for fuel assemblies of first, second, third and fourth year operation. (Authors)

  13. 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)

  14. Decommissioning of Brennilis NPP

    International Nuclear Information System (INIS)

    Baize, Jean-Marc

    1998-01-01

    This EDF press communique give information related to the decommissioning of the Brennilis NPP. The following five items are developed in this report: 1. the level-2 decommissioning operations at the Brennilis NPP; 2. the Brennilis NPP, a pilot operation from the commissioning up to the decommissioning; 3. history of the Brennilis NPP decommissioning; 4. the types of radioactive wastes generated by the Brennilis NPP decommissioning; 5. the Brennilis NPP - a yard management as a function of the wastes. The document contains also seven appendices addressing the following subjects: 1. the share of decommissioning assigned to EDF and the decommissioning steps; 2. the EDF installations in course of decommissioning; 3. the CEA decommissioned installations or in course of decommissioning; 4. regulations; 5. costs; 6. waste management - principles; 7. data on the decommissioning yard

  15. Dynamic model for tritium build-up at NPP with RBMK type reactors and its enviromental beraviour

    International Nuclear Information System (INIS)

    Badyaev, V.V.; Egorov, Yu.A.; Ivanov, E.A.; Stegachev, G.F.; Tolstykh, V.D.

    1982-01-01

    A model of tritium production dynamics for a high power NPP with RBMK type reactors is proposed and investigated. The main ''skeleton'' model structure for forecasting tritium buildup at a NPP and its exchange with the environment has been singled out at a heuristic level. Decomposition and layout of the units have been performed by global functional relations of the investigated objects (NPP and environment). the model accounts for only oxidized tritium forms. Water exchange between the NPP subsystems and environment is the main mechanism for tritium migration. The model does not account for scheduled periodic maintenance work effects, presence of stagnant zones in the station circuits, fuel burn-up, etc. The parametric identification method applied in the model makes the model adaptable to particular situations and considered systems of the NPP and environment. Completing the model with necessary and sufficient experimental data one can pass to certain forecasting problems and to NPP control as a tritium source in the environment

  16. NPP life management (abstracts)

    International Nuclear Information System (INIS)

    Litvinskij, L.L.; Barbashev, S.V.

    2002-01-01

    Abstracts of the papers presented at the International conference of the Ukrainian Nuclear Society 'NPP Life Management'. The following problems are considered: modernization of the NPP; NPP life management; waste and spent nuclear fuel management; decommissioning issues; control systems (including radiation and ecological control systems); information and control systems; legal and regulatory framework. State nuclear regulatory control; PR in nuclear power; training of personnel; economics of nuclear power engineering

  17. Burnup credit activities in the United States

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  18. Field experience of new nuclear fuel types on the Kola NPP

    International Nuclear Information System (INIS)

    Adeev, V.; Burlov, S.; Panov, A.; Saprykin, V.

    2008-01-01

    Specificity of the Kola nuclear power plant geographical position, conditions of region economics determine fuel management strategy. Isolation of Kola power supply system and, as a consequence, generating capacities redundancy cause operation of the nuclear power plant on reduced power level. At the same time there is a need to operate the power unit on the maximum power level in the case of not planned conditions. The basis of in-core fuel management is an achievement of the maximal burnup under providing of high installed capacity. At present there are not abilities to improve the fuel cycle based on traditional implementation fuel assemblies. Burnup maximum in these fuel cycles is achieved. At the core periphery installed highest possible quantity of the burned-up assemblies in the view of safety operation margins satisfaction. Works on application of the second generation fuel have been carried out on the Kola NPP since 2002. Fuel assemblies of this type are profiled. Burnable absorber, changed lattice spacing in relation to standard fuel, changed height of a fuel column, thickness of fuel pin clad are applied. In CR fuel followers modernized docking unit (with hafnium plates are intended for energy-release splash suppression) is used. At present 2-nd generation fuel is in experimental operation on unit 3 (18-21 fuel cycles, 2002-2007 years) and unit 4 (18-19 fuel cycles, 2005-2007 years). Safety margins did not exceeded. Coolant activity did not exceed the limiting value. There were not damaged fuel assemblies of second generation. Originally in the project of applications of new fuel it was supposed to refuel annually 78 fresh assemblies. At the moment annual refueling consists of 66 assemblies with effective enrichment 3.82 %. Cycle duration does not exceed 250-260 effective days. The part of assemblies is left on 5-th cycle of operation. In a similar fuel cycle in 2007 on the unit 1 operation with profiled fuel (enrichment of 3.82 %) of shakeproof type

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  20. Application of burnup credit concept to transport

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Nakagome, Yoshihiro.

    1994-01-01

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

  1. Phenomena and parameters important to burnup credit

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  2. Burnup credit for storage and transportation casks

    International Nuclear Information System (INIS)

    Wells, A.H.

    1988-01-01

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

  3. Technical development on burn-up credit for spent LWR fuels

    International Nuclear Information System (INIS)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

    2000-10-01

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

  4. 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.

  5. Technical development on burn-up credit for spent LWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

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

  6. Burnup determination of mass spectrometry for nuclear fuels

    International Nuclear Information System (INIS)

    Zhang Chunhua.

    1987-01-01

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

  7. Issues for effective implementation of burnup credit

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  8. Low-water considerations for NPP sites

    International Nuclear Information System (INIS)

    Jensen, J.; Frank, T.; Wahl, T.

    2009-01-01

    In the frame of reactor safety considerations flood protection and the evaluation of low-water events are safety relevant issues. Therefore low-water statistics were performed for the coastal region Cuxhaven and the Elbe river estuary. The consideration of the longitudinal profile of water levels during low-tide in connection with surface water effects, morphodynamic changes of the river and anthropogenic modifications is of importance for conclusions concerning the NPP sites in some distance of the reference tide gauge. The authors performed a statistical low-tide analysis for the NPP sites Brunsbuettel and Brokdorf.

  9. Nuclear fuel operation at Balakovo NPP

    International Nuclear Information System (INIS)

    Morozov, A.

    2015-01-01

    The presentation addressed the positive experience of the TVS-2M assemblies implementation at Balakovo NPP in 18 month fuel cycles, at uprated power (104%) and the usage of the axial profiled Gd-rods in order to minimize the power peaking factors and linear heat rate in the upper part in some of the fuel rods. The results of the test operation of fuel rods with different claddings, made by E110M, E125 and E635M alloys at Balakovo NPP were also provided. The recently observed problem with the “white crust” on the cladding surfaces was also discussed

  10. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

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

    2000-12-01

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

  11. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-01

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

  12. NPP service life management

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2001-01-01

    Problems of NPP service life management and service life prolongation are reviewed. Methods for the prolongation of the French NPP service life are discussed, priority directions of nuclear block service life management in regard to aging in the context of the European program of investigation into the materials aging are identified. Questions of the provision of the 60 years service life of the Mihama 1 block (Japan) and decision of the problem of the control equipment aging in Great Britain are discussed. Situation with the prolongation of licenses on the NPP operation in the USA and Spain is considered [ru

  13. Present status and future developments of the implementation of burnup credit in spent fuel management systems in Germany

    International Nuclear Information System (INIS)

    Neuber, J.C.

    1998-01-01

    The paper describes the experience gained in Germany in applying burnup credit methodologies to wet storage and dry transport systems of spent LWR fuel. It gives a survey of the levels of burnup credit presently used or intended to be used, the regulatory status and future developments planned, the codes used for performing depletion and criticality calculations, the methods applied to verification of these codes, and the methods used to treat parameters specific of burnup credit. In particular it is shown that the effect of axial burnup profiles on wet PWR storage designs based on burnup credit varies from fuel type to fuel type. For wet BWR storage systems the method of estimating a loading curve is described which provides for a given BWR fuel assembly design the minimum required initial burnable absorber content as a function of the initial enrichment of the fuel. (author)

  14. Influence of prolonged nuclear fuel burnup on safety characteristics of advanced PWRs

    International Nuclear Information System (INIS)

    Spasojevic, D.; Matausek, M.; Marinkovic, N.

    1989-01-01

    Prolonged nuclear fuel burnup in advanced NPP with four or more instead of three one-year cycles, and/or with 15- to 18-month instead of standard 12-month cycles, requires the fresh fuel to have increased enrichment combined with burnable poisons. This causes changes in axial and radial distribution of power generation during the particular fuel cycles, so that detailed analyses of thermal reliability of reactor core becomes necessary. This paper presents the results of the analysis of the departure from nuclear boiling ratio DNBR for an equilibrium cycle of an advanced PWR. (author)

  15. First steps towards modelling high burnup effect in UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    O` Carroll, C; Lassmann, K; Laar, J Van De; Walker, C T [CEC Joint Research Centre, Karlsruhe (Germany)

    1997-08-01

    High burnup initiates a process that can lead to major microstructural changes near the edge of the fuel: formation of subgrains, the loss of matrix fission gas and an increase in porosity. A consequence of this, is a decrease of thermal conductivity near the edge of the fuel which may be major implications for the performance of LWR fuels at higher burnup. The mechanism for the changes in grain structure, the apparent depletion of Xe and increase in porosity is associated with the high fission density at the fuel periphery. This is in turn due to the preferential capture of epithermal neutrons in the resonances of {sup 238}U. The new model TUBRNP predicts the radial burnup profile as a function of time together with the radial profile of plutonium. The model has been validated with data from LWR UO{sub 2} fuels with enrichments in the range 2 to 8.25% and burnups between 21 to 75 Gwd/t. It has been reported that at high burnup EPMA measures a sharp decrease in the concentration of Xe near the fuel surface. This loss of Xe is interpreted as a signal that the gas has been swept out of the original grains into pores: this ``missing`` Xe has been measured by XRF. It has been noted experimentally that the restructuring (Xe depletion and changes in grain structure) have an onset threshold local burnup in the region of 70 to 80 GWd/t: a specific value was taken for use in the model. For a given fuel TUBRNP predicts the local burnup profile, and the depth corresponding to the threshold value is taken to be the thickness of the Xe depleted region. The theoretical predictions have been compared with experimental data. The results are presented and should be seen as a first step in the development of a more detailed model of this phenomenon. (author). 22 refs, 9 figs, 2 tabs.

  16. Value of burnup credit beyond actinides

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  17. Burnup verification using the FORK measurement system

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1994-01-01

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

  18. Implementation of the KASKAD computer code system for WWER-440 at Kozloduy NPP

    International Nuclear Information System (INIS)

    Antonov, A.; Georgieva, N.; Spasova, V.

    2003-01-01

    Since 2002 at Kozloduy NPP - EP1 the code package KASKAD is used for WWER-440 reactor core calculations. The main codes entering this package are: BIPR-7A: 3-D diffusion and core analysis code; PERMAK-A: 2-D fine mesh diffusion code. The burnup calculations were performed for all cycles of the Kozloduy NPP Unit 1, Unit 2, Unit 3 and Unit 4. For the last 4-5 cycles of the Units were calculated control rods worth, critical boron concentration at zero power, reactivity coefficients and linear power. These results were analysed and were compared with experimental data. Some results were given in this paper

  19. Influence of fuel element burn-up on the power peaking factor in PWR; Vpliv zgorelosti gorivnega elementa na konicne faktorje moci v tlacnovodnem reaktorju

    Energy Technology Data Exchange (ETDEWEB)

    Ravnik, M; Mele, I [Institut ' Jozef Stefan' , Ljubljana (Yugoslavia); Falkowski, J [Institut energii atomowel, Swierk (Poland)

    1988-07-01

    Influence of fuel element burn-up distribution on radial power peaking factors is presented for Krsko NPP. The effect is strong for elements loaded in the periphery of the core with large power gradients. Neglecting the burn-up distributions inside fuel elements leads to {+-} 5% error on power peaking factor of the same element and {+-} 2% at other locations in the core. Influence on k is observed due to perturbed leakage from the core and due to redistribution of the importance function of the core. (author)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  2. Development of high burnup nuclear fuel technology

    International Nuclear Information System (INIS)

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

    1987-04-01

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

  3. Burnup analysis of the power reactor, 2

    International Nuclear Information System (INIS)

    Ezure, Hideo

    1975-09-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  5. Burnup calculation code system COMRAD96

    International Nuclear Information System (INIS)

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

    1997-06-01

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

  6. Burnup calculation code system COMRAD96

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

  7. NPP Krsko Containment Response Following Main Steam Line Break

    International Nuclear Information System (INIS)

    Spalj, S.; Grgic, D.; Cavlina, N.

    2000-01-01

    This paper presents the calculation of thermohydraulic environmental parameters (pressure and temperature) inside containment of Krsko NPP after postulated Main Steam Line Break (MSLB) accident. This analysis was done as a part of the ambient parameters specification in the frame of the NPP Krsko Equipment Qualification (EQ) project. The RELAP5/mod2 computer code was used for the determination of MSLB mass and energy release and computer code GOTHIC was used to calculate pressure and temperature profiles inside NPP Krsko containment. The analysis was performed for spectrum of break sizes to account for possible steam superheating during accidents with smaller break sizes. (author)

  8. 'CANDLE' burnup regime after LWR regime

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Nagata, Akito

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  11. Impact of the measurement data on the CORD-2 nuclear design calculations of the NPP Krsko

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2004-01-01

    The CORD-2 package was developed at Jozef Stefan Institute and has been validated for the nuclear design calculations of PWR cores. It has been used for the independent verification of the NPP Krsko nuclear design for the last 6 cycles of operation. The accuracy of the package is very good fulfilling all criteria usually imposed on the design prediction of the reactor nuclear parameters. To obtain as robust package as possible and to eliminate potential systematic errors of the package, it was decided to rely on measured core power distributions. In core power measurements, which are performed each month of reactor operation, are used to obtain fuel assemblies burnup histories. Consequently, burnup distributions obtained from the power measurements of all previous cycles are taken as a starting point at the beginning of the considered cycle. Since a lot of experience has been gained with the package, it was decided to evaluate the impact of measurement data on the accuracy of the calculations. Burnup calculations of all 19 cycles of the NPP Krsko are repeated, building simultaneously the calculated library of burnup histories for all fuel assemblies. The basic reactor parameters such as HZP critical boron concentration, isothermal temperature coefficient, control rod worth and cycle length are compared to the results obtained with CORD-2 standard sequence of calculation and direct measurements.(author)

  12. Systemization of burnup sensitivity analysis code. 2

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2005-02-01

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

  13. EMAS at Doel NPP

    International Nuclear Information System (INIS)

    Nolan, Peter; Thoelen, Els

    1998-01-01

    In October 1995, Doel NPP of Electrabel, Belgium opted to seek registration under the EC Eco-management and Audit Scheme (EMAS). A comprehensive environmental management system (EMS) has been introduced and implemented, encompassing all four PWRs and the supporting departments. A critical step was to seek certification from an accredited environmental auditing body against the International Standard ISO 14001. This provided the foundation for the publicly available environmental statement required by EMAS. The complications of achieving EMAS at a time when national and international standards were being re-formulated were successfully overcome and Doel NPP passed its EMAS audit in June 1997. (author)

  14. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    International Nuclear Information System (INIS)

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; Sterbentz, James W.

    2014-01-01

    Highlights: • The burnup of irradiated AGR-1 TRISO fuel was analyzed using gamma spectrometry. • The burnup of irradiated AGR-1 TRISO fuel was also analyzed using mass spectrometry. • Agreement between experimental results and neutron physics simulations was excellent. - Abstract: AGR-1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR-1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non-destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR-1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs-137 activity and the other based on the ratio of Cs-134 and Cs-137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA (fissions per initial heavy metal atom) for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can be determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP-MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma

  15. Experimental support of WWER-440 fuel reliability and serviceability at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, A; Ivanov, V; Pnyushkin, A [Nauchno-Issledovatel` skij Inst. Atomnykh Reaktorov, Dimitrovgrad (Russian Federation); Tzibulya, V [AO Mashinostroitelnij Zavod Electrostal (Russian Federation); Kolosovsky, V; Bibilashvili, Yu [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation); Dubrovin, K [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1994-12-31

    Results from post-reactor examination of two WWER-440 fuel assemblies spent at the Kola NPP Unit 3 during 4 and 5 fuel cycles are presented. The fuel assembly states and their serviceability allowance are estimated experimentally at the RIAR hot laboratory and studied by non-destructive and destructive methods. The following parameters are examined: fuel assembly overall dimensions change; fuel element diameter change; fuel element cladding corrosion and hydriding; fuel element cladding mechanical properties; fission gas release from fuel and gas pressure; fuel macro- and microstructure. it has been found that the maximum fuel burnup of fuel assemblies No. 1 and No.2 achieved is 58.3 and 64.0 MWd/kg, respectively. The mechanical fuel pellets-cladding interaction has been observed at the average fuel burnup above 45 MWd/kg that occurred with increasing the local cladding diameter at the areas of pellets end arrangement (bamboo stick). The gas release linearly increases at the range 2.7% per 10 MWd/kg within burnup of 43-60 MWd/kg. 9 figs., 3 refs.

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

    International Nuclear Information System (INIS)

    DOE

    1997-01-01

    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 eff , of a spent nuclear fuel package. Fifty-seven UO 2 , UO 2 /Gd 2 O 3 , and UO 2 /PuO 2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k eff (which can be a function of the trending parameters) such that the biased k eff , when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection

  17. 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

  18. Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

    International Nuclear Information System (INIS)

    Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V.; Boulyga, S.F.; Becker, J.S.

    2005-01-01

    An analytical method is described for the estimation of uranium concentrations, of 235 U/ 238 U and 236 U/ 238 U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10 -9 g/g to 2.0 x 10 -6 g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing 235 U/ 238 U and 236 U/ 238 U isotope ratios and the average value amounted to 9.4±0.3 MWd/(kg U). (orig.)

  19. Systemization of burnup sensitivity analysis code

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2004-02-01

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

  20. Temelin NPP commissioning experiences

    International Nuclear Information System (INIS)

    Hanus, V.

    2002-01-01

    The Building Permit for the Temelin NPP with four VVER units was issued in 1986, which is a long time ago. Since then, however, was taken a route that is very different from what anybody imagined. Described are the legislative and design changes and given is a current condition of the power plant

  1. Dukovany NPP maintenance management

    International Nuclear Information System (INIS)

    Siegel, F.

    2005-01-01

    Maintenance planning and management for the technological equipment of the Dukovany NPP are described. A Maintenance Control and Evaluation System has been developed and is in use to help manage the complex maintenance issue. Practical examples of outputs of the System, with a comprehensive use of the cost, reliability and safety related data, are presented. (author)

  2. NPP Krsko decommissioning concept

    International Nuclear Information System (INIS)

    Novsak, M.; Fink, K.; Spiler, J.

    1996-01-01

    At the end of the operational lifetime of a nuclear power plant (NPP) it is necessary to take measures for the decommissioning as stated in different international regulations and also in the national Slovenian law. Based on these requirements Slovenian authorities requested the development of a site specific decommissioning plan for the NPP KRSKO. In September 1995, the Nuklearna Elektrarna Krsko (NEK) developed a site specific scope and content for decommissioning plan including the assumptions for determination of the decommissioning costs. The NEK Decommissioning Plan contains sufficient information to fulfill decommissioning requirements identified by NRC, IAEA and OECD - NEA regulations. In this paper the activities and the results of development of NEK Decommissioning Plan consisting of the development of three decommissioning strategies for the NPP Krsko and selection of the most suitable strategy based on site specific, social, technical, radiological and economical aspects, cost estimates for the strategies including the costs for construction of final disposal facilities for fuel/high level waste (fuel/HLW) and low/intermediate level waste (LLW/ILW) and scheduling all activities necessary for the decommissioning of the NPP KRSKO are presented. (author)

  3. NPP Krsko decommissioning concept

    International Nuclear Information System (INIS)

    Novsak, M.; Fink, K.; Spiler, J.

    1996-01-01

    At the end of the operational lifetime of a nuclear power plant (NPP) it is necessary to take measures for the decommissioning as stated in different international regulations and also in the national Slovenian law. Based on these requirements Slovenian authorities requested the development of a site specific decommissioning plan for the NPP Krsko. In September 1995, the Nuklearna Elektrarna Krsko (NEK) developed a site specific scope and content for a decommissioning plan including the assumptions for determination of the decommissioning costs. The NEK Decommissioning Plan contains sufficient information to fulfill the decommissioning requirements identified by NRC, IAEA and OECD - NEA regulations. In this paper the activities and results of development of NEK Decommissioning Plan consisting of the development of three decommissioning strategies for the NPP Krsko and selection of the most suitable strategy based on site specific, social, technical, radiological and economic aspects, cost estimates for the strategies including the costs for construction of final disposal facilities for fuel/high level waste (fuel/HLW) and low/intermediate level waste (LLW/ILW) and scheduling of all activities necessary for the decommissioning of the NPP Krsko are presented. (author)

  4. Two managerial grids in NPP

    International Nuclear Information System (INIS)

    Zhao Hui

    2012-01-01

    Today, the nuclear power corporation (NPC) enjoys the profit of LCEP (the low carbon economic policy). at the same time, they also enduring more and more pressure. For example, the partner competition or the NPP potential occupational risk . The efficient counterplot of risk is the self-ability cultivation. It is essential to research the NPP managerial flow. The nuclear power plant (NPP) unit is a carrier of the NPC enterprise management system, and has taken on a new look 'pull one portion then the whole moving'. The NPP has three systematical characters, the security responsibility center, the man-machine system and the input-output system. The manufacturing system and the enterprise management system are the great constituents of the NPP managerial flows. Means of systems analysis, we can find out the truth of the NPP running interface. In CHINA, there are many operating experiences near 20 years. It indicates that the NPP manufacturing system and the enterprise system are the roots of the nuclear power corporation, the core of the all NPP systems must be based on it. So the ability cultivation is the work core to NPP. It is reliably to ensure the NPP to be up against problems, for instance, the security duty, the costing control and the man-machine system running harmoniously. This paper introduces the NPP managerial flow and the present state of QNPC, also come up with a proposal to refer for the NPC development actions of collective measure, specialization, standardization, fine. (author)

  5. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  6. Integrated burnup calculation code system SWAT

    International Nuclear Information System (INIS)

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

    1997-11-01

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

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

    International Nuclear Information System (INIS)

    2011-01-01

    compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the costs of code data library updating, the use of S/U analysis methodologies could be accomplished on a shorter schedule and a lower cost than the gathering of sufficient experimental data. ENERCON estimates of the costs of an updated S/U computer code and data suite are $5M to $10M with a schedule of two to three years. Recent ORNL analyses using the S/U analysis method show that the bias and uncertainty values for fission product cross sections are smaller than previously expected. This result is confirmed by a similar EPRI approach using different data and computer codes. ENERCON also found that some issues regarding the implementation of burnup credit appear to have been successfully resolved especially the axial burnup profile issue and the depletion parameter issue. These issues were resolved through data gathering activities at the Yucca Mountain Project and ORNL.

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

    Energy Technology Data Exchange (ETDEWEB)

    Enercon Services, Inc.

    2011-03-14

    compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the costs of code data library updating, the use of S/U analysis methodologies could be accomplished on a shorter schedule and a lower cost than the gathering of sufficient experimental data. ENERCON estimates of the costs of an updated S/U computer code and data suite are $5M to $10M with a schedule of two to three years. Recent ORNL analyses using the S/U analysis method show that the bias and uncertainty values for fission product cross sections are smaller than previously expected. This result is confirmed by a similar EPRI approach using different data and computer codes. ENERCON also found that some issues regarding the implementation of burnup credit appear to have been successfully resolved especially the axial burnup profile issue and the depletion parameter issue. These issues were resolved through data gathering activities at the Yucca Mountain Project and ORNL.

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

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

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

  10. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  11. Nuclear fuels with high burnup: safety requirements

    International Nuclear Information System (INIS)

    Phuc Tran Dai

    2016-01-01

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

  12. Burnup credit in a dry storage module

    International Nuclear Information System (INIS)

    Thornton, J.R.

    1989-01-01

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

  13. Status of burnup credit implementation in Switzerland

    International Nuclear Information System (INIS)

    Grimm, P.

    1998-01-01

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

  14. Advancements in reactor physics modelling methodology of Monte Carlo Burnup Code MCB dedicated to higher simulation fidelity of HTR cores

    International Nuclear Information System (INIS)

    Cetnar, Jerzy

    2014-01-01

    The recent development of MCB - Monte Carlo Continuous Energy Burn-up code is directed towards advanced description of modern reactors, including double heterogeneity structures that exist in HTR-s. In this, we exploit the advantages of MCB methodology in integrated approach, where physics, neutronics, burnup, reprocessing, non-stationary process modeling (control rod operation) and refined spatial modeling are carried in a single flow. This approach allows for implementations of advanced statistical options like analysis of error propagation, perturbation in time domain, sensitivity and source convergence analyses. It includes statistical analysis of burnup process, emitted particle collection, thermal-hydraulic coupling, automatic power profile calculations, advanced procedures of burnup step normalization and enhanced post processing capabilities. (author)

  15. Influence of high burnup on the decay heat power of spent fuel at long-term storage

    International Nuclear Information System (INIS)

    Bergelson, B.; Gerasimov, A.; Tikhomirov, G.

    2005-01-01

    Development and application of advanced fuel with higher burnup is now in practice of NPP with light water reactors in an increasing number of countries. High burnup allows to decrease significantly consumption of uranium. However, spent fuel of this type contains increased amount of high active actinides and fission products in comparison with spent fuel of common-type burnup. Therefore extended time of storage, improved cooling system of the storage facility will be required along with more strong radiation protection during storage, transportation and processing. Calculated data on decay heat power of spent uranium fuel of light water VVER-1000 type reactor are discussed in the paper. Long-term storage of discharged fuel during 100000 years is considered. Calculations were made for burnups of 40-70 MW d/kg. In the initial 50-year period of storage, power of fission products is much higher than that of actinides. Power of gamma-radiation is mainly due to fission products. During subsequent storage power of fission products quickly decreases, the main contribution to the power is given by actinides rather than by fission products. (author)

  16. The octopus burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

  17. The OCTOPUS burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

  18. The octopus burnup and criticality code system

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  19. The OCTOPUS burnup and criticality code system

    International Nuclear Information System (INIS)

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

    1996-06-01

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

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

    International Nuclear Information System (INIS)

    Ewing, R.I.

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  2. Burnup calculation for a tokamak commercial hybrid reactor

    International Nuclear Information System (INIS)

    Feng Kaiming; Xie Zhongyou

    1990-08-01

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

  3. Optimum burnup of BAEC TRIGA research reactor

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  4. 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.

  5. Burnup credit effect on proposed cask payloads

    International Nuclear Information System (INIS)

    Hall, I.K.

    1989-01-01

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

  6. WWER-1000 Burnup Credit Benchmark (CB5)

    International Nuclear Information System (INIS)

    Manolova, M.A.

    2002-01-01

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

  7. Whole core burnup calculations using 'MCNP'

    International Nuclear Information System (INIS)

    Haran, O.; Shaham, Y.

    1996-01-01

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

  8. Whole core burnup calculations using `MCNP`

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  9. Application of routine methods for the inspector fuel burn-up determination and identification of displacement of spent fuel elements by dummy elements

    International Nuclear Information System (INIS)

    Rohar, S.

    1979-08-01

    14 irradiated assemblies were analyzed using nondestructive high resolution gamma spectrometry (HRGS). Measured and calculated (on the basis of calorimetric data) axial burnup profiles and average burnup values were compared. The measurements of spent fuel were performed in the Bohunice A-1 dry hot cell by using a proper collimating system and the standard Agency equipment, consisting of PGT intrinsic Ge detectors and Silena MCA with 1024 channels. The method of 134 Cs/ 137 Cs fission product activity ratio was used for burnup determination. It was found that the burnup values for 14 measured assemblies determined by HRGS were systematically lower than the calculated values with about 4-5%. The difference between the nondestructively determined burnup value of the 2N0053 assembly (average over 11 measured points) and destructively determined burnup (average over 19 measured points) was less than 2%. Passive neutron measurements of the irradiated assembly showed that the neutron counting rate was high enough for practical use and that the neutron and gamma profiles were similar and close to the burnup profile. Some calculations of gamma ray activity angular distribution were made for different numbers of dummy elements inside the irradiated assemblies. The results show that, by using gamma spectrometry transversal method, it is possible to find a significant number of dummy elements in different types of assemblies

  10. Burnup calculations using Monte Carlo method

    International Nuclear Information System (INIS)

    Ghosh, Biplab; Degweker, S.B.

    2009-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Iqbal Masood

    2008-01-01

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

  12. Nuclear fuel utilization at the Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Boyadzhiev, Z [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Kharalampieva, Ts; Pejchinov, Ts

    1996-12-31

    Data on core loading and operation conditions during past fuel cycles of the Units 1 to 6 at the Kozloduy NPP are presented. The Units 1 and 2 have reached average discharge fuel burn-up of 31 MW d/kg U, the Unit 3 - 34 MW d/kg U and the Unit 4 - 36.5 MW d/kg U. By use of dummies and low-leakage core loading patterns for WWER-440 cores an effective reduction in reactor pressure vessel irradiation is obtained. By increasing the enrichment level and improving the characteristics of the Units 3 and 4, a design fuel cycles duration has been reduced by 5-10% in number of assemblies. Core loading design has been modelled using computer codes SPPS-1, BIPR-7, ALBOM, PROROC. A 3-year fuel cycle utilizing 4.4% enriched fuel proved to be more efficient for WWER-1000 by 15% reduction in fuel cost compared to the 2-year cycle. Future developments include improvements of the in-core monitoring system and process on-line simulation based on more accurate computer codes. 7 refs., 6 figs., 10 tabs.

  13. New-Generation NASA Aura Ozone Monitoring Instrument (OMI) Volcanic SO2 Dataset: Algorithm Description, Initial Results, and Continuation with the Suomi-NPP Ozone Mapping and Profiler Suite (OMPS)

    Science.gov (United States)

    Li, Can; Krotkov, Nickolay A.; Carn, Simon; Zhang, Yan; Spurr, Robert J. D.; Joiner, Joanna

    2017-01-01

    coarser spatial and spectral resolution of the Suomi National Polar-orbiting Partnership (Suomi-NPP) Ozone Mapping and Profiler Suite (OMPS) instrument, application of the new PCA algorithm to OMPS data produces highly consistent retrievals between OMI and OMPS. The new PCA algorithm is therefore capable of continuing the volcanic SO2 data record well into the future using current and future hyperspectral UV satellite instruments.

  14. Burnup credit activities being conducted in the United States

    International Nuclear Information System (INIS)

    Lake, W.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Kato, Tomoaki; Yamashita, Kiyonobu

    2007-01-01

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

  16. Automated generation of burnup chain for reactor analysis applications

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  17. Automated generation of burnup chain for reactor analysis applications

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

  18. Dukovany NPP operation

    International Nuclear Information System (INIS)

    Vlcek, Jaroslav

    2010-01-01

    The topics discussed include: Dukovany NPP among CEZ Group power plants; International missions at the plant; Plant operation results; and Strategic goals and challenges. Historical data are presented in the graphical form, such as the unit capacity factor, unplanned capability loss factor, unplanned automatic scrams, fuel reliability, industrial safety accident rate, collective radiation exposure, WANO index, power generation data, and maximum achievable power by the end of year. Also discussed were the company culture and human resources, maintenance, power uprate, and related phenomena. (P.A.)

  19. Cernavoda NPP Knowledge Transfer

    International Nuclear Information System (INIS)

    Valache, C. M.

    2016-01-01

    Full text: The paper presents a description of the Knowledge Transfer (KT) process implemented at Cernavoda NPP, its designing and implementation. It is underlined that applying a KT approach should improve the value of existing processes of the organization through: • Identifying business, operational and safety risks due to knowledge gaps, • Transfer of knowledge from the ageing workforce to the peers and/or the organization, • Continually learning from successes and failures of individual or teams, • Convert tacit knowledge to explicit knowledge, • Improving operational and safety performance through creating both new knowledge and better access to existing knowledge. (author

  20. Operation safety at Ignalina NPP

    International Nuclear Information System (INIS)

    Zheltobriukh, G.

    1999-01-01

    An improvement of operational safety at Ignalina NPP covers: improvement of management structure and safety culture; symptom-based emergency operating procedures; staff training and full scope simulator; program of components ageing; metal inspection; improvement of fire safety. The first plan of Ignalina NPP Safety culture development for 1997 purposed to the SAR recommendation implementation was prepared and approved by the General Director

  1. Preparation for Ignalina NPP decommissioning

    International Nuclear Information System (INIS)

    Medeliene, D.

    2004-01-01

    Latest developments of atomic energy in Lithuania, works done to prepare Ignalina NPP for final shutdown and decommissioning are described. Information on decommissioning program for Ignalina NPP unit 1, decommissioning method, stages and funding is presented. Other topics: radiation protection, radioactive waste management and disposal. Key facts related to nuclear energy in Lithuania are listed

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

    International Nuclear Information System (INIS)

    Chernushev, V.; Sokolov, F.

    2002-01-01

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

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

    International Nuclear Information System (INIS)

    Esteve, B.; Gueldner, R.; Hoffman, R.; Watteau, M.

    2002-01-01

    In every sense of the term, nuclear fuel forms the core of nuclear power plants (NPPs). Although there are many equipment items important for their safety function or for their participation in NPP availability, the fuel, in essence renewable, is one of the key elements which have to be acted upon if utilities are to be helped to fulfil their mission of generating power in total safety and supplying the kWh to their customers at the best price. Nuclear fuel is also the core business of the Framatome-ANP Fuel Business Group: pooling and rationalising the available skills - technical, cultural and human - supplied by each of the partners forms a challenge which it is up to each and every one to meet in a cooperative spirit. This paper gives an outline of the company's extended burnup experience, current R and D, and its plans for the future. (author)

  4. Fission product margin in burnup credit analyses

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  5. Perspectives of Living PSA in NPP Krsko

    International Nuclear Information System (INIS)

    Vrbanic, I.; Kastelan, M.

    1996-01-01

    Nuclear power plant Krsko has completed the Level 1/Level 2 Probabilistic Safety Analysis (PSA) for internal initiating events and is in the process of completing the same for the external initiators. The analysis completed up to now has provided a valuable insight into a plant risk profile. In NPP Krsko there is a plan to use the PSA model as a permanent tool for the risk based applications and incorporate it into a decision making process. In order to achieve this there is a need to permanently maintain the PSA model in a manner that it reflects both the plan configuration/design at a time point and the operational experience up to the time point. All the activities aimed toward keeping the PSA model up-to-dated in this sense are usually referred to as a Living PSA (LPSA) program. NPP Krsko is in the process of defining and proceduralizing a LPSA program that would be plant specific and based on known world practices. Further, in order to be suitable for risk based applications the PSA model must be flexible in a sense that modifications to the base case model may be done easily and requantifications performed quickly as to evaluate various conditions imposed by real or hypothetical situations. NPP Krsko PSA model has been based on licensing type software. The requirements specified above dictate the transfer of the overall model to an application oriented software of newer generation with larger capabilities. The transfer becomes a part of a mentioned ongoing effort aimed at establishing LPSA model and concept. The paper present this effort and the perspectives of LPSA concept and risk based applications in NPP Krsko. (author)

  6. Study of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    Pavelescu, M.; Borza, M.

    1975-01-01

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

  7. Operational benchmark for VVER-1000, unit 6, Kozloduy NPP

    International Nuclear Information System (INIS)

    Apostolov, T.; Petrov, B.

    1999-01-01

    Benchmark calculations have been carried out using the 3D nodal code TRAPEZ. Global neutron-physics characteristics of the VVER-1000 core, Kozloduy NPP Unit 6, have been determined taking into account the real loading patterns and operational history of the first three cycles. The code TRLOAD has been used to perform the fuel reloading between any two cycles. The reactor and components descriptions as well as material compositions are given. The results presented include the critical boric acid concentration, the radial power distribution, the axial power distribution for the maximum overload assembly, and the burnup distribution at three different moments during each cycle. Calculated values have been compared with measured data. It is shown that the results obtained by the TRAPEZ code are in good agreement with the experimental data. The information presented could serve as a test case for validation of code packages designed for analyzing the steady-state operation of VVERs. (author)

  8. Criticality safety analysis of the NPP Krsko storage racks

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2002-01-01

    NPP Krsko is going to increase the capacity of the spent fuel storage pool by replacement of the existing racks with high-density racks. This will be the second reracking campaign since 1983 when storage was increased from 180 to 828 storage locations. The pool capacity will increase from 828 to 1694 with partial reracking by the spring 2003. The installed capacity will be sufficient for the current design plant lifetime. Complete reracking of the spent fuel pool will additionally increase capacity to 2321 storage locations. The design, rack manufacturing and installation has been awarded to the Framatome ANP GmbH. Burnup credit methodology, which was approved by the Slovenian Nuclear Safety Administration in previous licensing of existing racks, will be again implemented in the licensing process with the recent methodology improvements. Specific steps of the criticality safety analysis and representative results are presented in the paper.(author)

  9. Determination of source term for Krsko NPP extended fuel cycle

    International Nuclear Information System (INIS)

    Nemec, T.; Persic, A.; Zagar, T.; Zefran, B.

    2004-01-01

    The activity and composition of the potential radioactive releases (source term) is important in the decision making about off-site emergency measures in case of a release into environment. Power uprate of Krsko NPP during modernization in 2000 as well as changing of the fuel type and the core design have influenced the source term value. In 2003 a project of 'Jozef Stefan' Institute and Slovenian nuclear safety administration determined a plantspecific source term for new conditions of fuel type and burnup for extended fuel cycle. Calculations of activity and isotopic composition of the core have been performed with ORIGEN-ARP program. Results showed that the core activity for extended 15 months fuel cycle is slightly lower than for the 12 months cycles, mainly due to larger share of fresh fuel. (author)

  10. Conceptual cask design with burnup credit

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  11. WWER-type NPP spray ponds screen

    International Nuclear Information System (INIS)

    Nikolova, M.; Jordanov, M.; Denev, J.; Markov, D.

    2003-01-01

    The objective of this study is to develop a protection screen of WWER-type NPP spray ponds. The screen design is to ensure reduction of the water droplets blown by the wind and, if possible, their return back to the spray ponds. The cooling capacity of the ponds is not to be changed below the design level for safety reasons. Computational fluid dynamics analysis is used to assess the influence of each design variant on the behavior of the water droplets distribution. Two variants are presented here. The one with plants is found not feasible. The second variant, with steel screen and terrain profile modification is selected for implementation. (author)

  12. Evaluation of rod insertion issue for NPP Krsko

    International Nuclear Information System (INIS)

    Gunstek, A.; Kurincic, B.

    1998-01-01

    The last couple of years incident with control rods sticking in lower part of the fuel assemblies have been reported of several reactor operators and fuel vendors throughout of the world. Several activities were initiated immediately to determine the root cause of incomplete rod insertion. The purpose of this activities were to collect plants trip history data and testing results, review of available worldwide experience, review of plant operation and fuel management, detailed review of manufacturing and material property and to maintain detailed mechanical model. In this paper, we will present activities in Nuclear Power Plant Krsko which have been performed after NRC initiated the Root Cause Process (NRC Bulletin 96-01). NPP Krsko has not experienced rod insertion anomaly yet but anyway the additional tests were carried out. Rod drop time measurements that were performed normally at beginning of cycle at nominal temperature and pressure (HSB mode) have been extended also to end of cycle. Rod drop time, velocity of dropped rods and magnitudes of the initial recoil bounces vs. burnup were also analyzed. Also RCCA drag test with upper internals in place and drive shafts attached to RCCAs has been performed since then. At last two outages (1997 and 1998) drag test were carried out with digital scale meter to gather additional information. In addition to that, the reload core design has been performed with new constrains on rodded fuel assembly burnup as proposed by the industry.(author)

  13. Fuel utilization experience in Bohunice NPP and regulatory requirements for implementation of progressive fuel management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Patenyi, V [Nuclear Regulatory Authority, Bratislava (Slovakia); Darilek, P; Majercik, J [Vyskumny Ustav Jadrovych Elektrarni, Trnava (Slovakia)

    1994-12-31

    The experience gained in fuel utilization and the basic requirements for fuel licensing in the Slovak NPPs is described. The original project of WWER-440 reactors supposes 3-year fuel cycle with cycle length of about 320 full power days (FPD). Since 1984 it was reduced to 290 FPD. Based on the experience of other countries, a 4-year fuel cycle utilization started in 1987. It is illustrated with data from the Bohunice NPP units. Among 504 fuel assemblies left for the fourth burnup cycle no leakage was observed. The mean burnup achieved in the different units varied from 33.1 to 38.5 Mwd/kg U. The new fuel assemblies used are different from the recent ones in construction, thermohydraulics, water-uranium ratio, enrichment and material design. To meet the safety criteria, regulatory requirements for exploitation of new fuel in WWER-440 were formulated by the Nuclear Regulatory Authority of Slovak Republic. 1 tab., 5 refs.

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

    International Nuclear Information System (INIS)

    Dufek, Jan; Valtavirta, Ville

    2014-01-01

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

  15. A Monte Carlo burnup code linking MCNP and REBUS

    International Nuclear Information System (INIS)

    Hanan, N. A.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1993-01-01

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

  17. Implementation of burnup credit in spent fuel management systems

    International Nuclear Information System (INIS)

    Dyck, H.P.

    2001-01-01

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

  18. NPP Prevlaka - Preparation of construction

    International Nuclear Information System (INIS)

    Bojic, K.

    1984-01-01

    On the basis of study 'Optimal electricity generation structure till the year 2000' production of 3 x 500 MWe in nuclear power plants has been anticipated. Second Croatian-Slovenian NPP project will be based on the same principles the first one (NPP Krsko) was based on. Preconstruction investigation studies are performed at site Prevlaka on river Sava downstream of Zagreb. Licensing procedure has started with republic Urban countryside planning activities. Preconstruction activities are planned to be finished by the end of 1986. while the construction is expected to start during 1987. Parallel to investigation studies for NPP Prevlaka, evaluation of nuclear technology and reactor type is planned to be made. (author)

  19. The seismic reassessment Mochovce NPP

    International Nuclear Information System (INIS)

    Baumeister, P.

    2004-01-01

    The design of Mochovce NPP was based on the Novo-Voronez type WWER-440/213 reactor - twin units. Seismic characteristic of this region is characterized by very low activity. Mochovce NPP site is located on the rock soil with volcanic layer (andesit). Seismic reassessment of Mochovce NPP was done in two steps: deterministic approach up to commissioning confirmed value Horizontal Peak Ground Acceleration HPGA=0.1 g and activities after commissioning as a consequence of the IAEA mission indicate higher hazard values. (author)

  20. Technology for NPP decantate treatment realized at Kola NPP

    International Nuclear Information System (INIS)

    Stakhiv, Michael; Avezniyazov, Slava; Savkin, Alexander; Fedorov, Denis; Dmitriev, Sergei; Kornev, Vladimir

    2007-01-01

    At Moscow SIA 'Radon' jointly with JSC 'Alliance Gamma', the technology for NPP Decantate Treatment was developed, tested and realized at Kola NPP. This technology consists of dissolving the salt residue and subsequent treatment by ozonization, separation of the deposits formed from ozonization and selective cleaning by ferro-cyanide sorbents. The nonactive salt solution goes to an industrial waste disposal site or a repository specially developed at NPP sites for 'exempt waste' products by IAEA classification. This technology was realized at Kola NPP in December 2006 year. At this time more than 1000 m 3 of decantates log time stored are treated. It allows solving very old problem to empty decantates' tanks at NPPs in environmentally safe manner and with high volume reduction factor. (authors)

  1. NPOESS Preparatory Project (NPP) Science Overview

    Science.gov (United States)

    Butler, James J.

    2011-01-01

    NPP Instruments are: (1) well understood thanks to instrument comprehensive test, characterization and calibration programs. (2) Government team ready for October 25 launch followed by instrument activation and Intensive Calibration/Validation (ICV). NPP Data Products preliminary work includes: (1) JPSS Center for Satellite Applications and Research (STAR) team ready to support NPP ICV and operational data products. (2) NASA NPP science team ready to support NPP ICV and EOS data continuity.

  2. Workflow in Almaraz NPP

    International Nuclear Information System (INIS)

    Gonzalez Crego, E.; Martin Lopez-Suevos, C.

    2000-01-01

    Almaraz NPP decided to incorporate Workflow into its information system in response to the need to provide exhaustive follow-up and monitoring of each phase of the different procedures it manages. Oracle's Workflow was chosen for this purpose and it was integrated with previously developed applications. The objectives to be met in the incorporation of Workflow were as follows: Strict monitoring of procedures and processes. Detection of bottlenecks in the flow of information. Notification of those affected by pending tasks. Flexible allocation of tasks to user groups. Improved monitoring of management procedures. Improved communication. Similarly, special care was taken to: Integrate workflow processes with existing control panels. Synchronize workflow with installation procedures. Ensure that the system reflects use of paper forms. At present the Corrective Maintenance Request module is being operated using Workflow and the Work Orders and Notice of Order modules are about to follow suit. (Author)

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

    International Nuclear Information System (INIS)

    Zhang Jian; Yu Hong; Gang Zhi

    2012-01-01

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

  4. Burnup calculation in microcells of high conversion reactors

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Jurkevicius, A.; Remeikis, V.

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-12-01

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

  7. Shock absorber in Ignalina NPP

    International Nuclear Information System (INIS)

    Bulavas, A.; Muralis, J.

    1996-09-01

    Theoretical calculation and experimental analysis of models of shock absorber in Ignalina NPP is presented. The results obtained from the investigation with model of shock absorber coincide with the theoretical calculation. (author). 2 figs., 3 refs

  8. Dukovany NPP - Safely 16 TERA

    International Nuclear Information System (INIS)

    Vlcek, J.

    2008-01-01

    In this presentation increasing of power output of the Dukovany NPP is reviewed. To operate all Dukovany Units safely with the perspective of long-term operation (LTO) of 50 - 60 years it is proposed.

  9. Selection of NPP for Kazakhstan

    International Nuclear Information System (INIS)

    Zhotabaev, Zh.R.

    2003-01-01

    Commercial NPP for Kazakhstan should to meet to several main requirements: 1). Safety operation (accident probability not more than 10 -6 1/p. year). 2). High efficiency > 40 %. 3). Possibility of use for high-temperature chemistry and hydrogen production. 4). Possibility for manufacturing of considerable part of equipment in Kazakhstan. 5). Possibility for fuel production and reprocessing in Kazakhstan. 6). Independence from existence of large water-supply sources. Comparative analysis of several NPP with different reactors (WWR-1000, Candu, BREST, VG-400; graphite molten salt reactor) shows that NPP with the graphite molten salt reactor meets to all above requirements, but hydrogen production it is possible by more complete 4-stage technology, since coolant temperature is 800 Deg. C. The principle advantage is possibility of manufacturing of main equipment and fuel in Kazakhstan that reduce the cost of NPP construction and operation

  10. Modernization programme at Dukovany NPP

    International Nuclear Information System (INIS)

    Trnka, M.

    2000-01-01

    The main goal of each NPP is to produce electricity safely, economically and without influence to environment. For Dukovany NPP it means to upgrade all documentation and perform the Equipment Upgrading Programme. All these activities are time and money consuming and therefore the determination of priority of all items was necessary. In the presentation there are mentioned some important changes in documentation, results of PSA studies and reason for Equipment Upgrading Programme performance. It was selected the most important item from the list of Equipment Upgrading Programme the I and C upgrading. Management has decided that Dukovany NPP will become among the best NPPs with WWER type of reactor. It seems this decision is the best way how to extend lifetime of the NPP. (author)

  11. Analyses for licensing of new fuel types at Paks NPP

    International Nuclear Information System (INIS)

    Kereszturi, A.; Bogatyr, S.; Miko, S.; Nemes, I.

    2003-01-01

    In the last years Paks NPP initiated several projects aiming at the introduction of new fuel types and resulting in more economic fuel cycles. The motivations, the reasons, and the economic consequences of the above modifications are detailed. The application of a new fuel type requires the renewal of the relevant chapters of the Safety Analysis Report. The fulfilment of fuel design basis requirements, to be summarised briefly also in the paper, must be investigated during normal and accidental conditions. The characteristics of the different codes, the data transfer between them are detailed. After, the cases of the Normal Operation, Anticipated Operation Occurrence, and the Postulated Accidents, judged as the most relevant ones in case of fuel modifications, are overviewed. In the last part, selected examples of the licensing calculations, performed by the above tools are presented. In conclusion, modifications of the WWER fuel, namely increased enrichment, application of burnable fuel pins, modified geometry make more economic fuel cycles (larger discharge burnup, power up-rate, reduced pressure vessel fluence) are possible. The further step (increased enrichment, burnable poison) of the fuel modernisation at NPP Paks is necessary for more economic fuel cycles and fuel consuming. A sound basis of licensing methodology, safety analysis, and necessary computer codes for the WWER fuel modernisation is available

  12. Burnup measurements of leader fuel elements

    International Nuclear Information System (INIS)

    Henriquez, C; Navarro, G; Pereda, C

    2000-01-01

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

  13. Psychology of NPP operation safety

    International Nuclear Information System (INIS)

    Tret'yakov, V.P.

    1993-01-01

    The book is devoted to psychologic investigations into different aspects of NPP operative personnel activities. The whole set of conditions on which successful and accident-free personnel operation depends, is analysed. Based on original engineering and socio-psychologic investigations complex psychologic support for NPP personnel and a system of training and upkeep of operative personnel skills are developed. The methods proposed have undergone a practical examination and proved their efficiency. 154 refs., 12 figs., 9 tabs

  14. Burnup characteristics of binary breeder reactors

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  15. Determination of nuclear fuel burn-up

    International Nuclear Information System (INIS)

    Kristak, J.; Vobecky, M.

    1973-01-01

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

  16. 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.

  17. Validation of the CORD-2 System for the Nuclear Design Calculations of the NPP Krsko Core

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2016-01-01

    The CORD-2 package intended for core design calculations of PWRs has be recently updated with some improved models. Since the modifications could substantially influence the obtained results, a technical validation process is required. This paper presents comparison of some calculated and measured parameters of the NPP Krsko core needed to qualify the package. Critical boron concentrations at hot full power for selected cycle burnup points and several parameters obtained during the start-up testing at the beginning of each cycle (hot zero power critical concentration, isothermal temperature coefficient and rods worth) for all 27 finished cycles of operation are considered. In addition, assembly-wise power distribution for some selected cycles is checked. Comparison has shown very good agreement of the CORD-2 calculated values with the selected measured parameter of the NPP Krsko core.(author).

  18. Technological and licensing challenges for high burnup fuel

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

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

  20. Establishing a PWR burn-up library

    International Nuclear Information System (INIS)

    Lutz, D.C.

    1981-01-01

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

  1. Device for measuring a burnup degree

    International Nuclear Information System (INIS)

    Ito, Toshiaki; Goto, Seiichiro

    1979-01-01

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

  2. Actinides burnup in a sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-15

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

  3. ABB PWR fuel design for high burnup

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  4. NPP Decommissioning: the concept; state of activities

    International Nuclear Information System (INIS)

    Nemytov, S.; Zimin, V.

    2001-01-01

    The main principles of NPP decommissioning concept in Russia are given. The conditions with fulfillment of works on NPP unit pre-decommissioning and decommissioning including: development of the normative documentation, creation of special fund for financing NPP decommissioning activities, deriving the Gosatomnadzor license for decommissioning of shut down NPP units, development of the equipment and technologies for waste and spent fuel management are presented. The decommissioning cost and labour intensity of one WWER-440 unit are shown. The practical works, executed on shut down units at Beloyarsk NPP (Unit1 and 2) and Novo Voronezh NPP (Unit 1 and 2) are outlined

  5. 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)

  6. Benchmarking burnup reconstruction methods for dynamically operated research reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

  7. Revised SWAT. The integrated burnup calculation code system

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide

    2000-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  10. Automated generation of burnup chain for reactor analysis applications

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  11. Probabilistic assessment of dry transport with burnup credit

    International Nuclear Information System (INIS)

    Lake, W.H.

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Boshoven, J.K.

    1992-09-01

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

  13. A Monte Carlo burnup code linking MCNP and REBUS

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-02-01

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

  15. Spain in South Ukraine NPP

    International Nuclear Information System (INIS)

    Ibanez, M.

    1994-01-01

    A Technical Assistance Protocol was signed between the Governments of the GIS and the Commission of the European Union (CEU) on August 2, 1991 and this was the starting point of the TACIS program. In this article, the activities described are those related to the TACIS-92/93/94 on site technical assistance to South Ukraine NPP (SUK NPP). Within the scope of the TACIS 92 Program the CEU and the Ukrainian Authorities agreed a list of projects to be implemented at South Ukraine NPP with the aim to improve the operational safety of the plant. This part of the program is called TACIS 92 on-site activities. The total budget allocated to these projects is a MECU. The European Union ''utility'' selected to lead this program at South Ukraine NPP was UNESA and the first contract to cover our activities was signed in July 1993 between the CEU (Mr. Pablo Benavides) and UNESA (Mr. Pedro Rivero). The projects will be implemented at SUK NPP but according to the contract UNESA is ''The Consultant'' and GOSKOMATON (The Ukrainian Sate Committee on Nuclear Power Utilization) is the ''Recipient Institution''. (Author)

  16. Emergency preparedness at Ignalina NPP

    International Nuclear Information System (INIS)

    Kairys, A.

    1998-01-01

    Brief review of Ignalina NPP safety upgrading and personnel preparedness to act in cases of accidents is presented. Though great activities are performed in enhancing the plant operation safety, the Ignalina NPP management pays a lot of attention to preparedness for emergency elimination and take measures to stop emergency spreading. A new Ignalina NPP emergency preparedness plan was drawn up and became operational. It is the main document to carry out organizational, technical, medical, evacuation and other activities to protect plant personnel, population, the plant and the environment from accident consequences. Great assistance was rendered by Swedish experts in drawing this new emergency preparedness plan. The plan consists of 3 parts: general part, operative part and appendixes. The plan is applied to the Ignalina NPP personnel, Special and Fire Brigade and also to other contractor organizations personnel carrying out works at Ignalina NPP. There are set the following emergency classes: incident, emergency situation, alert, local emergency, general emergency. Separate intervention level corresponds to each emergency class. Overview of personnel training to act in case of an emergency is also presented

  17. Regulatory approach to NPP ageing in Bulgaria

    International Nuclear Information System (INIS)

    Vassilev, D.

    2000-01-01

    In this contribution summary information of Kozloduy NPP units is presented. The nuclear legislation, regulatory approach for managing safety aspects on NPP ageing, short term programme, complex programme PRG'97 ant other aspects of ageing management are discussed

  18. Some problems of NPP personnel training

    International Nuclear Information System (INIS)

    Vajshnis, P.P.; Kumkov, L.P.; Omel'chuk, V.V.

    1984-01-01

    Shortcomings of NPP personnel training are discussed. Development of full-scale training systems is necessary for qualitative training operative personnel. Primary problems that should be necessarily solved for ensuring effective training NPP personnel are considered

  19. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

  20. HAMCIND, Cell Burnup with Fission Products Poisoning

    International Nuclear Information System (INIS)

    Abe, Alfredo Y.; Dos Santos, Adimir

    2002-01-01

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

  1. The commercial impact of burnup increase

    International Nuclear Information System (INIS)

    Fenzlein, C.; Schricker, W.

    2002-01-01

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

  2. Regulatory aspects of NPP safety

    International Nuclear Information System (INIS)

    Kastchiev, G.

    1999-01-01

    Extensive review of the NPP Safety is presented including tasks of Ministry of Health, Ministry of Internal Affairs, Ministry of Environment and Waters, Ministry of Defense in the field of national system for monitoring the nuclear power. In the frame of national nuclear safety legislation Bulgaria is in the process of approximation of the national legislation to that of EC. Detailed analysis of the status of regulatory body, its functions, organisation structure, responsibilities and future tasks is included. Basis for establishing the system of regulatory inspections and safety enforcement as well as intensification of inspections is described. Assessment of safety modifications is concerned with complex program for reconstruction of Units 1-4 of Kozloduy NPP, as well as for modernisation of Units 5 and 6. Qualification and licensing of the NPP personnel, Year 2000 problem, priorities and the need of international assistance are mentioned

  3. Individual protection of NPP personnel

    International Nuclear Information System (INIS)

    Koshcheev, V.S.; Gol'dshtejn, D.S.; Chetverikova, Z.S.

    1983-01-01

    Specific features of NPP personnel individual protection are considered, mainly with respect to maintenance and repair works on various type reactors. The major concern is given to the selection and application reglamentations of the individual protection system (IPS), employment of sanitary locks, the organization of individual protection under the conditions of a heating microclimate. The ways are specified to the development and introduction of the most effective IPS and improvement of the entire NPP personnel individual protection system with respect to providing the necessary protection effect for maintaining high working capability of the personnel and minimizing the IPS impact on human organism functional systems. The accumulated experience in the personnel individual protection can be applied during construction and operation of NPP's in CMEA member-countries [ru

  4. Safety culture at Mochovce NPP

    International Nuclear Information System (INIS)

    Markus, Jozef; Feik, Karol

    2002-01-01

    This article presents the approach of Mochovce NPP to the Safety culture. It presents activities, which have been taken by Mochovce NPP up to date in the area of Safety culture enhancement with the aim of getting the term into the subconscious of each employee, and thus minimising the human factor impact on occurrence of operational events in all safety areas. The article furthermore presents the most essential information on how the elements characterising a continuous progress in reaching the planned Safety culture goals of the company management have been implemented at Mochovce NPP, as well as the management's efforts to get among the best nuclear power plant operators in this area and to be an example for the others. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  6. Experimental programmes related to high burnup fuel

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  7. High burnup models in computer code fair

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  8. High burnup models in computer code fair

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  9. Radioactive wastes management of NPP

    International Nuclear Information System (INIS)

    Klyuchnikov, A.A.; Pazukhin, Eh.M.; Shigera, Yu. M.; Shigera, V.Yu.

    2005-01-01

    Modern knowledge in the field of radiation waste management on example of the most serious man-made accident at Chernobyl NPP are illuminated. This nuclear power plant that after accident in 1986 became in definite aspect an experimental scientific ground, includes all variety of problems which have to be solved by NPP personnel and specialists from scientific organizations. This book is aimed for large sphere of readers. It will be useful for students, engineers, specialists and those working in the field of nuclear power, ionizing source and radiation technology use for acquiring modern experience in nuclear material management

  10. NPP construction cost in Canada

    International Nuclear Information System (INIS)

    Gorshkov, A.L.

    1988-01-01

    The structure of capital costs during NPP construction in Canada is considered. Capital costs comprise direct costs (cost of the ground and ground rights, infrastructure, reactor equipment, turbogenerators, electrotechnical equipment, auxiliary equipment), indirect costs (construction equipment and services, engineering works and management services, insurance payments, freight, training, operating expenditures), capital per cents for the period of construction and cost of heavy water storages. It proceeds from the analysis of the construction cost structure for a NPP with the CANDU reactor of unit power of 515, 740 and 880 MW, that direct costs make up on the average 62%

  11. NPP Krsko secondary side analysis

    International Nuclear Information System (INIS)

    Fabijan, Lj.

    1987-01-01

    The purpose of this work is to analyze secondary side thermohydraulics response on steam generator tube plugging in order to ensure nominal NPP power. We had established that the additional opening of the governing valve No. 3 and 4 can compensate pressure drop caused by steam generator tube plugging. Two main steam flows with four governing valves were simulated. Steam expansion in turbine and feed water system was modeled separately. All important process point and steam moisture changes impact on nominal NPP power were analysed. (author)

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

    International Nuclear Information System (INIS)

    Jaschik, W.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi

    2005-11-01

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

  14. Highlights on R and D work related to the achievement of high burnup with MOX fuel in commercial reactors

    International Nuclear Information System (INIS)

    Lippens, M.; Maldague, Th.; Basselier, J.; Boulanger, D.; Mertens, L.

    2000-01-01

    Part of the R and D work made at BELGONUCLEAIRE in the field of high burnup achievement with MOX fuel in commercial LWRs is made through lnternational Programmes. Special attention is given to the evolution with burnup of fuel neutronic characteristics and of in-reactor rod thermal-mechanical behaviour. Pu burning in MOX is characterized essentially by a drop of Pu 239 content. The other Pu isotopes have an almost unchanged concentration, due to internal breeding. The reactivity drop of MOX versus burnup is consequently much less pronounced than in UO 2 fuel. Concentration of minor actinides Am and Cm becomes significant with burnup increase. These nuclides start to play a role on total reactivity and in the helium production. The thermal-mechanical behaviour of MOX fuel rod is very similar to that of UO 2 . Some specificities are noticed. The better PCI resistance recognized to MOX fuel has recently been confirmed. Three PWR MOX segments pm-irradiated up to 58 GWd/tM were ramped at 100 W/cm.min respectively to 430-450-500 W/cm followed by a hold time of 24 hours. No segment failed. MOX and UO 2 fuels have different reactivities and operate thus at different powers. Moreover, radial distribution of power in MOX pellet is less depressed at high burnup than in UO 2 , leading to higher fuel central temperature for a same rating. The thermal conductivity of MOX fuel decreases with Pu content, typically 4% for 10% Pu. The combination of these three elements (power level, power profile, and conductivity) lead to larger FGR at high burnup compared to UO 2 . Helium production remains low compared to fission gas production (ratio < 0.2). As faster diffusing element, the helium fractional release is much higher than that of fission gas, leading to rod pressure increase comparable to the one resulting from fission gas. (author)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu

    2017-06-01

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

  17. Burnup credit demands for spent fuel management in Ukraine

    International Nuclear Information System (INIS)

    Medun, V.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Leppaenen, J.; Isotalo, A.

    2012-01-01

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

  19. A microcomputer program for coupled cycle burnup calculations

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Thomas, W.

    1992-01-01

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

  1. Two dimensional burn-up calculation of TRIGA core

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  3. The V-1 NPP and V-2 NPP upgrading

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities in the V-1 NPP and V-2 NPP upgrading as well as maintenance carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. The V-1 NPP applied the so called 'Small Backfitting Programme'covering 81 points of the Czechoslovak Atomic energy Commission Decree No 5/91. Continual upgrading continued after the Backfitting Programme completion with the Safety Report and following Nuclear Regulatory Authority of Slovak Republic (NRA SR) Decrees No 1/94 and 110/94 setting spheres and procedure for adopting and implementation of measures enabling the units to operate further on. Results of expert missions, analyses and assessments of components identified by Basic Engineering became the basis for the development of the Gradual Reconstruction Programme. The Programme outputs underwent economic and probabilistic assessing their contribution to nuclear safety. This process resulted in finalizing the Gradual Reconstruction Programme which started to be implemented in 1996 and will be completed in 1999. It is implemented by the REKON consortium and covers 17 areas including Instrumentation and Control, self-consumption emergency supply, leakage monitoring, emergency core cooling system, seismic reinforcement and radioactivity localisation. Both units will reach internationally acceptable safety standards for the remaining life-time period. The V-2 NPP Upgrading and Safety Enhancement Programme includes results of activities performed in the course of last years to define all important activities leading to enhancement of nuclear safety and performance reliability and effectiveness within the plant life-time period and to establish conditions for extending the life-time of these units for 40 years. The V-2 NPP Upgrading and Safety Enhancement Programme aims to assure safe operation with a probability of the core damages less than 10 -4 /reactor · year

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  5. Safety aspects of NPP ageing

    International Nuclear Information System (INIS)

    1995-01-01

    Preparation of safety practices on assessment and management of aging of major NPP components important to safety, CRP on management of aging of concrete containment buildings, CRP on management of aging of in-containment instrumentation and control cables are outlined

  6. Qualification of NPP operations personnel

    International Nuclear Information System (INIS)

    Wang Jiao.

    1987-01-01

    Competence of personnel is one of the important problems for safety operation of nuclear power plant. This paper gives a description of some aspects, such as the administration of NPP, posts, competence of personnel, training, assessing the competence and personnel management

  7. 3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Lake, W.H.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Gerhards, E.

    1993-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

  11. Parametric neutronic analyses related to burnup credit cask design

    International Nuclear Information System (INIS)

    Parks, C.V.

    1989-01-01

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

  12. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

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

  14. TRIGA criticality experiment for testing burn-up calculations

    International Nuclear Information System (INIS)

    Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz

    1999-01-01

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

  15. An optimal model for fuel burnup in nuclear reactors

    International Nuclear Information System (INIS)

    Anton, V.

    1979-05-01

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

  16. Impact of extended burnup on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-04-01

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

  17. 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)

  18. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

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

    2010-02-01

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

  19. 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)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  1. Application of Candle burnup to small fast reactor

    International Nuclear Information System (INIS)

    Sekimoto, H.; Satoshi, T.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Dalle, Hugo M.

    2009-01-01

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

  3. Optimum Discharge Burnup and Cycle Length for PWRs

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. Fission gas release at high burn-up: beyond the standard diffusion model

    International Nuclear Information System (INIS)

    Landskron, H.; Sontheimer, F.; Billaux, M.R.

    2002-01-01

    At high burn-up standard diffusion models describing the release of fission gases from nuclear fuel must be extended to describe the experimental loss of xenon observed in the fuel matrix of the rim zone. Marked improvements of the prediction of integral fission gas release of fuel rods as well as of radial fission gas profiles in fuel pellets are achieved by using a saturation concept to describe fission gas behaviour not only in the pellet rim but also as an additional fission gas path in the whole pellet. (author)

  5. Burnup degree measuring device for spent fuel

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  6. Study on the sensitivity of Self-Powered Neutron Detectors (SPND) and its change due to burn-up

    International Nuclear Information System (INIS)

    Cho, Gyuseong; Lee, Wanno; Yoon, Jeong-Hyoun.

    1996-01-01

    Self-Powered Neutron Detectors (SPND) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. While they have several advantages such as small size, low cost, and relatively simple electronics required in conjunction with its usage, it has some intrinsic problems of the low level of output current, a slow response time, the rapid change of sensitivity which makes it difficult to use for a long term. In this paper, Monte Carlo simulation was accomplished to calculate the escape probability as a function of the birth position for the typical geometry of rhodium-based SPNDs. Using the simulation result, the burn-up profile of rhodium number density and the neutron sensitivity is calculated as a function of burn-up time in the reactor. The sensitivity of the SPND decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long-term usage. (author)

  7. Profiles

    International Nuclear Information System (INIS)

    2004-01-01

    Profiles is a synthetic overview of more than 100 national energy markets in the world, providing insightful facts and key energy statistics. A Profile is structured around 6 main items and completed by key statistics: Ministries, public agencies, energy policy are concerned; main companies in the oil, gas, electricity and coal sectors, status, shareholders; reserve, production, imports and exports, electricity and refining capacities; deregulation of prices, subsidies, taxes; consumption trends by sector, energy market shares; main energy projects, production and consumption prospects. Statistical Profiles are present in about 3 pages the main data and indicators on oil, gas, coal and electricity. (A.L.B.)

  8. Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

    Energy Technology Data Exchange (ETDEWEB)

    Mironov, V.P.; Matusevich, J.L.; Kudrjashov, V.P.; Ananich, P.I.; Zhuravkov, V.V. [Inst. of Radiobiology, Minsk Univ. (Belarus); Boulyga, S.F. [Inst. of Inorganic Chemistry and Analytical Chemistry, Johannes Gutenberg-Univ. Mainz, Mainz (Germany); Becker, J.S. [Central Div. of Analytical Chemistry, Research Centre Juelich, Juelich (Germany)

    2005-07-01

    An analytical method is described for the estimation of uranium concentrations, of {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10 cm soil layers at the investigated sampling sites varied from 2.1 x 10{sup -9}g/g to 2.0 x 10{sup -6}g/g depending mainly on the distance from Chernobyl NPP. A slight variation of the degree of burn-up of spent reactor uranium was revealed by analyzing {sup 235}U/{sup 238}U and {sup 236}U/{sup 238}U isotope ratios and the average value amounted to 9.4{+-}0.3 MWd/(kg U). (orig.)

  9. Characteristics of WWER-1000 fuel rod claddings and FA components from E635 alloy at burnups up to 72 MWd/kgU

    International Nuclear Information System (INIS)

    Nikulin, A.; Novikov, A.; Peregud, M.; Shishov, V.; Shevyakov, A.; Volkova, I.; Novoselov, A.; Kobylyansky, G.

    2011-01-01

    In this paper operation experience, results of investigated E365 alloy components of Balakovo NPP Unit 1 and Kalinin NPP unit 1 fuel assemblies are presented. Appearance, shape changes and geometric size, corrosion state of guide thimbles, angles and fuel rods, corrosion of fuel claddings are studied. At the end authors concluded that: I) E635 alloy corroborated its high operation reliability as fuel claddings and WWER-1000 FA components during 6 year service to the fuel burnup of 72MWd/kgU; II) Based on the results from the post-irradiation investigations of the fuel rods and other structural elements of WWER-1000 FAA, fabricated from E635 alloy, in terms of the basic operational characteristics, their resources after the 6 year operation cycle have not been exhausted; III) The geometrical parameters, corrosion states, tensile properties of items fabricated from fuel alloy did not attain the values that would prevent their further operation: 1) the elongations of the fuel rods at the mean burnups up to 66.2 MWd/kgU do not exceed 15 mm or 4.9%; 8) the amount of the oxide coat at surface of GT and CT does not exceed 45 μm, the hydrogen content is <0.03% mass; 9) the oxide coat at the surfaces of the frame angles does not exceed 50 μm, the hydrogen content is <0.04% mass

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Tatsumi, Masahiro; Hyoudou, Hideaki

    2008-08-01

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

  12. ASSET experience at Paks NPP

    International Nuclear Information System (INIS)

    Szabo, I.

    1997-01-01

    At Paks NPP special attention has been paid to international reviews since the very beginning of operation. Several international teams visited Paks in order to provide independent assessment of plant performance, conditions and safety. Paks NPP Management has the further intention to invite international reviews regularly (yearly) in the future as well. The experience gained during these reviews helped to establish a unified process of preparation for the reviews, performing them and handling the results. The Safety Department is in charge of organization of the whole process. All these reviews have their specific features and they are focused on different areas. The ASSET reviews provides the assessment of plant performance and safety through the analysis of safety significant events, which have occurred at the nuclear power plant. This approach makes this review specific and different from the other ones

  13. Retrofitting of NPP Computer systems

    International Nuclear Information System (INIS)

    Pettersen, G.

    1994-01-01

    Retrofitting of nuclear power plant control rooms is a continuing process for most utilities. This involves introducing and/or extending computer-based solutions for surveillance and control as well as improving the human-computer interface. The paper describes typical requirements when retrofitting NPP process computer systems, and focuses on the activities of Institute for energieteknikk, OECD Halden Reactor project with respect to such retrofitting, using examples from actual delivery projects. In particular, a project carried out for Forsmarksverket in Sweden comprising upgrade of the operator system in the control rooms of units 1 and 2 is described. As many of the problems of retrofitting NPP process computer systems are similar to such work in other kinds of process industries, an example from a non-nuclear application area is also given

  14. ASSET experience at Paks NPP

    Energy Technology Data Exchange (ETDEWEB)

    Szabo, I [Operational Safety Dept., Paks NPP, Paks (Hungary)

    1997-10-01

    At Paks NPP special attention has been paid to international reviews since the very beginning of operation. Several international teams visited Paks in order to provide independent assessment of plant performance, conditions and safety. Paks NPP Management has the further intention to invite international reviews regularly (yearly) in the future as well. The experience gained during these reviews helped to establish a unified process of preparation for the reviews, performing them and handling the results. The Safety Department is in charge of organization of the whole process. All these reviews have their specific features and they are focused on different areas. The ASSET reviews provides the assessment of plant performance and safety through the analysis of safety significant events, which have occurred at the nuclear power plant. This approach makes this review specific and different from the other ones.

  15. High Frequency Acoustic Microscopy for the Determination of Porosity and Young's Modulus in High Burnup Uranium Dioxide Nuclear Fuel

    Science.gov (United States)

    Marchetti, Mara; Laux, Didier; Cappia, Fabiola; Laurie, M.; Van Uffelen, P.; Rondinella, V. V.; Wiss, T.; Despaux, G.

    2016-06-01

    During irradiation UO2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of both porosity and elastic properties in high burnup UO2 pellet can be investigated via high frequency acoustic microscopy. For this purpose ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A UO2 pellet with a burnup of 67 GWd/tU was characterized using the acoustic microscope installed in the hot cells of the JRC-ITU at a 90 MHz frequency, with methanol as coupling liquid. VR was measured at different radial positions. A good agreement was found, when comparing the porosity values obtained via acoustic microscopy with those determined using SEM image analysis, especially in the areas close to the centre. In addition, Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile and to the hardness radial profile data obtained by Vickers micro-indentation.

  16. Nuclear Fuel in Cofrentes NPP

    International Nuclear Information System (INIS)

    2002-01-01

    Fuel is an essential in the nuclear power generating business because of its direct implications on safety, generating costs and the operating conditions and limitations of the facility. Fuel management in Cofrentes NPP has been targeted at optimized operation, enhanced reliability and the search for an in-depth knowledge of the design and licensing processes that will provide Iberdrola,as the responsible operator, with access to independent control of safety aspects related to fuel and free access to manufacturing markets. (Author)

  17. NPP Krsko Living PSA Concept

    International Nuclear Information System (INIS)

    Vrbanic, I.; Spiler, J.

    2000-01-01

    NPP Krsko developed PSA model of internal and external initiators within the frame of the Individual Plant Examination (IPE) project. Within this project PSA model was used to examine the existing plant design features. In order to continue with use of this PSA model upon the completion of IPE in various risk-informed applications in support of plant operation and evaluations of design changes, an appropriate living PSA concept needed to be defined. The Living PSA concept is in NPP Krsko considered as being a set of activities pursued in order to update existing PSA model in a manner that it appropriately represents the plant design, operation practice and history. Only a PSA model which is being updated in this manner can serve as a platform for plant-specific risk informed applications. The NPP Krsko living PSA concept is based on the following major ponts. First, the baseline PSA model is defined, which is to be maintained and updated and which is to be reference point for any risk-informed application. Second, issues having a potential for impact on baseline PSA model are identified and procedure and responsibilities for their permanent monitoring and evaluation are established. Third, manner is defined in which consequential changes to baseline PSA model are implemented and controlled, together with associated responsibilities. Finally, the process is defined by which the existing version of baseline PSA model is superseded by a new one. Each time a new version of baseline PSA model is released, it would be re-quantified and the results evaluated and interpreted. By documenting these re-quantifications and evaluations of results in a sequence, the track is being kept of changes in long-term averaged risk perspective, represented by long-term averaged frequencies of core damage and pre-defined release categories. These major topics of NPP Krsko living PSA concept are presented and discussed in the paper. (author)

  18. Welding facilities for NPP assembling

    International Nuclear Information System (INIS)

    Rojtenberg, S.S.

    1987-01-01

    Recommendations concerning the choice of equipment for welding in pre-assembling work shops, in the enlarging assembling shops and at the assembling site, are given. Advanced production automatic welders and semiautomatic machines, applied during the NPP equipment assembling as well as automatic machines specially produced for welding the main reactor components and pipelines are described. Automatic and semiautomatic machine and manual welding post supply sources are considered

  19. Performance of the NPP CrIS Sensor

    Science.gov (United States)

    Emch, P. G.; Farrow, S. V.; Gu, D.; Wang, C.; Hagan, D. E.; Sabet-Peyman, F.

    2009-12-01

    This paper discusses the performance of the first Cross-track Infrared Sounder (CrIS) flight instrument and its data product performance. Together with ATMS, the CrIS sensor is a critical payload for National Polar-orbiting Operational Environmental Satellite System (NPOESS) providing temperature and moisture profiles, and will first fly on the NPOESS Preparatory Project (NPP) mission, the risk reduction flight for NPOESS. NPOESS is the next generation weather and climate monitoring system for the Department of Defense and National Oceanic and Atmospheric Administration (NOAA), being developed under contract by Northrop Grumman Aerospace Systems. The NPP flight sensor has recently completed thermal vacuum, electro-magnetic interference, and vibration testing. By the time of the presentation, the sensor sell-off is expected to have been completed as well. An overview of the sensor, the FM1 measurement performance, and details of the retrieval algorithms will be provided in this presentation.

  20. New approach of second Romanian NPP siting

    International Nuclear Information System (INIS)

    Mauna, Traian

    2010-01-01

    The NPP sitting studies in Romania began before 1975. The first Romanian NPP CANDU 6 type reactor gone to erection in 1980 on Cernavoda site planned to have 5 units. Gained the experience from Cernavoda NPP sitting, the first mission of new multi-branch of specialists team was to choose new NPP sites adapting the NPP Cernavoda project to the new parameters of close water cooling circuit and hard less and no rock foundation strata. The studies were carrying out in different stages on the inner rivers Olt, Mures, Somes in Transylvania historical region. This paper tries to reconsider shortly the old analysis according to the last IAEA Safety Standards, taking into account the new NPP generation requirement. Paper is focused on geological aspects and other local sites characteristics. (authors)

  1. Licensing of the Ignalina NPP

    International Nuclear Information System (INIS)

    Kutas, S.

    1999-01-01

    Since 1991 State Nuclear Power Safety Inspectorate (VATESI) has regulated Ignalina NPP operation by issuing annual operating permits. Those have been issued following submission of specified documents by the Ignalina NPP that have been reviewed by VATESI. However, according to to the procedures that are now established in the Law on Nuclear Energy and subordinate regulations the use of nuclear energy in the Republic of Lithuania is subject to strict licensing. Therefore a decision about the licence for continued operation of unit 1 should be taken. Licence would be granted by VATESI in cooperation with the Ministry of Health, Ministry of Environment and the institutions of local authorities. Ignalina NPP presented to the VATESI safety analysis report (SAR) with other documents. SAR was made mainly by foreign experts and financed by European Bank for Reconstruction and Development (EBRD). VATESI in this process is supported by western regulators. A special project LAP - Licensing Assistance Project was launched to help VATESI perform licensing according western practices

  2. On the NPP structural reliability

    International Nuclear Information System (INIS)

    Klemin, A.I.; Polyakov, E.F.

    1980-01-01

    Reviewed are the main statements peculiarities and possibilities of the first branch guiding technical material (GTM) ''The methods of calculation of structural reliability of NPP and its systems at the stage of projecting''. It is stated, that in GTM presented are recomendations on the calculation of reliability of such specific systems, as the system of the reactor control and protection the system of measuring instruments and automatics and safe systems. GTM are based on analytical methods of modern theory of realibility with the Use of metodology of minimal cross sections of complex systems. It is stressed, that the calculations on the proposed methods permit to calculate a wide complex of reliability parameters, reflecting separately or together prorerties of NPP dependability and maintainability. For NPP, operating by a variable schedule of leading, aditionally considered are parameters, characterizing reliability with account of the proposed regime of power change, i.e. taking into account failures, caused by decrease of the obtained power lower, than the reguired or increase of the required power higher, than the obtained

  3. Metamorphosis of NPP A1, V1, V2

    International Nuclear Information System (INIS)

    Dobak, D.; Moncekova, M.

    2005-01-01

    In this book the history of construction, commissioning and exploitation of NPP A1, NPP V1 and NPP V2 in Jaslovske Bohunice is presented on documentary photos. Vicinity around of these NPPs is presented, too

  4. Ignalina NPP: living and working conditions

    International Nuclear Information System (INIS)

    Chiuzhas, A.

    1998-01-01

    The conference was devoted to discuss the social problems related with the operation of Ignalina NPP. The main topics are the following: analysis of public opinion of surrounding region of Ignalina NPP including neighbouring Daugavpils district in Latvia, environment impact evaluation of Daugavpils district, assessment of the influence of Ignalina NPP operation to the development of business in the region, investigation of problems of Visaginas town - residence of Ignalina NPP personnel. The specificity of Visaginas (former Sniechkus) is defined by the majority of non-native Lithuanians living there. Cultural transformation and political organization of the region were surveyed as well

  5. Corrosion behavior of duplex and reference cladding in NPP Grohnde

    International Nuclear Information System (INIS)

    Besch, O.A.; Yagnik, S.K.; Eucken, C.M.; Bradley, E.R.

    1996-01-01

    The Nuclear Fuel Industry Research (NFIR) Group undertook a lead test assembly (LTA) program in NPP Grohnde PWR in Germany to assess the corrosion performance of duplex and reference cladding. Two identical 16 by 16 LTAs, each containing 32 peripheral test rods, completed four reactor cycles, reaching a peak rod burnup of 46 MWd/kgU. The results from poolside examinations performed at the end of each cycle, together with power histories and coolant chemistry, are reported. Five different cladding materials were characterized during fabrication. The corrosion performance of the cladding materials was tracked in long-term tests in high-pressure, high-temperature autoclaves. The relative ranking of corrosion behavior in such tests corresponded well with the in-reactor corrosion performance. The extent and distribution of hydriding in duplex and reference specimens during the autoclave testing has been characterized. The in-reactor corrosion data indicate that the low-tin Zircaloy-4 reference cladding, R2, had an improved corrosion resistance compared to high-tin Zircaloy-4 reference cladding, R1. Two types of duplex cladding, D1 (Zr-2.5% Nb) and D2 (Zr-0.4% Fe-0.5% Sn), showed an even further improvement in corrosion resistance compared to R2 cladding. The third duplex cladding, D3 (Zr-4 + 1.0% Nb), had significantly less corrosion resistance, which was inferior to R1. The in-reactor and out-reactor corrosion performances have been ranked

  6. Fuel burnup analysis for the Moroccan TRIGA research reactor

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  8. Use of burnup credit for transportation and storage

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  9. Status of burnup credit implementation and research in Switzerland

    International Nuclear Information System (INIS)

    Grimm, P.

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  14. TRIGA fuel element burnup determination by measurement and calculation

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  15. Modelling of some high burnup phenomena in nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  17. Fuel cycle cost considerations of increased discharge burnups

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  20. Fission gas release from fuels at high burnup

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  1. Increased fuel burn-up and fuel cycle equilibrium

    International Nuclear Information System (INIS)

    Debes, M.

    2001-01-01

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

  2. COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. Role of measurement systems in burnup credit operations

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  4. Burnup credit implementation plan and preparation work at JAERI

    International Nuclear Information System (INIS)

    Nomura, Y.; Itahara, K.

    2001-01-01

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

  5. Optimization of TRU burnup in modular helium reactor

    International Nuclear Information System (INIS)

    Yonghee, Kim; Venneri, F.

    2007-01-01

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

  6. Radiation monitoring in the NPP environment, control of radioactivity in NPP-environment system

    International Nuclear Information System (INIS)

    Egorov, Yu.A.

    1987-01-01

    Problems of radiation monitoring and control of the NPP-environment system (NPPES) are considered. Radiation control system at the NPP and in the environment provides for the control of the NPP, considered as the source of radioactive releases in the environment and for the environmental radiation climate control. It is shown, that the radiation control of the NPP-environment system must be based on the ecological normalization principles of the NPP environmental impacts. Ecological normalization should be individual for the NPP region of each ecosystem. The necessity to organize and conduct radiation ecological monitoring in the NPP regions is pointed out. Radiation ecological monitoring will provide for both environmental current radiation control and information for mathematical models, used in the NPPES radiation control

  7. Consequences of the increase of burnup on the fuel

    International Nuclear Information System (INIS)

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

    1986-04-01

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

  8. Fission gas release from fuel at high burnup

    International Nuclear Information System (INIS)

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

    1978-03-01

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

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

    International Nuclear Information System (INIS)

    Gyori, G.

    2011-01-01

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

  10. Burnup measurements with the Los Alamos fork detector

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  11. Isotopic biases for actinide-only burnup credit

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Kidman, R.B.

    1976-01-01

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

  13. Findings of an international study on burnup credit

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  15. A regime showing anomalous triton burnup in JET

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  16. Display systems for NPP control

    International Nuclear Information System (INIS)

    Rozov, S.S.

    1988-01-01

    Main trends in development of display systems used as the means for image displaying in NPP control systems are considered. It is shown that colour display devices appear to be the most universal means for concentrated data presentation. Along with digital means the display systems provide for high-speed response, sufficient for operative control of executive mechanisms. A conclusion is drawn that further development of display systems will move towards creation of large colour fields (on reflection base or with multicolour gas-discharge elements)

  17. Emergency preparedness at Barsebaeck NPP in Sweden

    International Nuclear Information System (INIS)

    Olsson, R.; Lindvall, C.

    1998-01-01

    On-site emergency preparedness plan at Barsebaeck NPP is presented. In an emergency the responsibility of the NPP is to alarm the emergency organizations, spend all efforts to restore safe operation, assess the potential source term as to size and time, protect their own personnel, inform personnel and public. Detailed emergency procedures overview is provided

  18. Safety culture in Ignalina NPP, regulatory view

    Energy Technology Data Exchange (ETDEWEB)

    Maksimovas, G [VATESI (Lithuania)

    1997-09-01

    The presentation describes how success on the way to a high level Safety Culture in Ignalina NPP may be achieved by daily, well motivated activities with good attitude and proper management participation, ensuring the development and proper implementation of Safety Culture principles within the activities of Operational organization of Ignalina NPP.

  19. Safety culture in Ignalina NPP, regulatory view

    International Nuclear Information System (INIS)

    Maksimovas, G.

    1997-01-01

    The presentation describes how success on the way to a high level Safety Culture in Ignalina NPP may be achieved by daily, well motivated activities with good attitude and proper management participation, ensuring the development and proper implementation of Safety Culture principles within the activities of Operational organization of Ignalina NPP

  20. Constructive approaches to the space NPP designing

    International Nuclear Information System (INIS)

    Eremin, A.G.; Korobkov, L.S.; Matveev, A.V.; Trukhanov, Yu.L.; Pyshko, A.P.

    2000-01-01

    An example of designing a space NPP intended for power supply of telecommunication satellite is considered. It is shown that the designing approach based on the introduction of a leading criterion and dividing the design problems in two independent groups (reactor with radiation shield and equipment module) permits to develop the optimal design of a space NPP [ru

  1. NPP Bohunice experience with ASSET services

    International Nuclear Information System (INIS)

    Klimo, J.

    1996-01-01

    The general description of Bohunice NPP ASSET experience history was given at the last annual workshop in 1995. In my short presentation I would like to pay attention to the progress in this area which was achieved at our NPP during the last year. (author)

  2. NPP Bohunice experience with ASSET services

    Energy Technology Data Exchange (ETDEWEB)

    Klimo, J [Bohunice NPP (Slovakia)

    1997-12-31

    The general description of Bohunice NPP ASSET experience history was given at the last annual workshop in 1995. In my short presentation I would like to pay attention to the progress in this area which was achieved at our NPP during the last year. (author).

  3. Bohunice V1 NPP upgrading programme

    International Nuclear Information System (INIS)

    Kerak, J.

    2001-01-01

    The paper describes whole process of Bohunice V1 NPP nuclear safety and operational reliability level increase which has been performed since units commissioning (1. unit in 1978, 2. unit in 1980), continued Small Reconstruction (1991 -1993) and finished Gradual Upgrading(1994 -2000). The main purpose is to last stage -Gradual upgrading of Bohunice V1 NPP. (author)

  4. Chernobyl NPP accident. Overcoming experience. Acquired lessons

    International Nuclear Information System (INIS)

    Nosovskij, A.V.; Vasil'chenko, V.N.; Klyuchnikov, A.A.; Prister, B.S.

    2006-01-01

    This book is devoted to the 20 anniversary of accident on the Chernobyl NPP unit 4. History of construction, causes of the accident and its consequences, actions for its mitigation are described. Modern situation with Chernobyl NPP decommissioning and transferring of 'Ukryttya' shelter into ecologically safe system are mentioned. The future of Chernobyl site and exclusion zone was discussed

  5. Seismic characterization of the NPP Krsko site

    International Nuclear Information System (INIS)

    Obreza, J.

    2000-01-01

    The goal of NPP Krsko PSA Project Update was the inclusion of plant changes (i.e. configuration/operational related) through the period January 1, 1993 till the OUTAGE99 (April 1999) into the integrated Internal/External Level 1/Level 2 NPP Krsko PSA RISK SPECTRUM model. NPP Krsko is located on seismotectonic plate. Highest earthquake was recorded in 1917 with magnitude 5.8 at a distance of 7-9 km. Site (founded) on Pliocene sediments which are as deep as several hundred meters. No surface faulting at the Krsko site has been observed and thus it is not to be expected. NPP Krsko is equipped with seismic instrumentation, which allows it to complete OBE (SSE). The seismic PSA successfully showed high seismic margin at Krsko plant. NPP Krsko seismic design is based on US regulations and standards

  6. Influence of FGR complexity modelling on the practical results in gas pressure calculation of selected fuel elements from Dukovany NPP

    International Nuclear Information System (INIS)

    Lahodova, M.

    2001-01-01

    A modernization fuel system and advanced fuel for operation up to the high burnup are used in present time in Dukovany NPP. Reloading of the cores are evaluated using computer codes for thermomechanical behavior of the most loaded fuel rods. The paper presents results of parametric calculations performed by the NRI Rez integral code PIN, version 2000 (PIN2k) to assess influence of fission gas release modelling complexity on achieved results. The representative Dukovany NPP fuel rod irradiation history data are used and two cases of fuel parameter variables (soft and hard) are chosen for the comparison. Involved FGR models where the GASREL diffusion model developed in the NRI Rez plc and standard Weisman model that is recommended in the previous version of the PIN integral code. FGR calculation by PIN2k with GASREL model represents more realistic results than standard Weisman's model. Results for linear power, fuel centre temperature, FGR and gas pressure versus burnup are given for two fuel rods

  7. Operating Experience at NPP Krsko

    International Nuclear Information System (INIS)

    Kavsek, D.; Bach, B.

    1998-01-01

    Systematic analysis of operational experience by assessment of internal and industry events and the feedback of lessons learned is one of the essential activities in the improvement of the operational safety and reliability of the nuclear power plant. At NPP Krsko we have developed a document called ''Operating Experience Assessment Program''. Its purpose is to establish administrative guidance for the processing of operating events including on-site and industry events. Assessment of internal events is based on the following methods: Event and Causal Factor Charting, Change Analysis, Barrier Analysis, MORT (Management Oversight and Risk Tree Analysis) and Human Performance Evaluation. The operating experience group has developed a sophisticated program entitled ''Operating experience tracking system'' (OETS) in response to the need for a more efficient way of processing internal and industry operating experience information. The Operating Experience Tracking System is used to initiate and track operational events including recommended actions follow up. Six screens of the system contain diverse essential information which allows tracking of operational events and enables different kinds of browsing. OETS is a part of the NPP Krsko nuclear network system and can be easily accessed by all plant personnel. (author)

  8. Analytical and experimental justification of safe operation of fuel loads of VVER reactors at Rovno NPP

    International Nuclear Information System (INIS)

    Andrianov, A.; Zagrebelny, L.

    2011-01-01

    The main task during the nuclear fuel operation us ensuring of the larger fuel burnup and as a result - reduction of the fuel constituent in the electricity production cost. The neutron-physic calculations are performed using the qualified codes BIPR-7A and PERMAK developed by Kurchatov Institute. The limits for calculated parameters are set by the Ukrainian regulations. Calculation results are documented in reports subject for independent expert review requested by the regulatory authority. In this report the following item have presented: 1) metrological check and calibration of measuring channels; 2) fuel cycles at Rivne NPP; 3) determination of experimental values of thermal-physic and neutron-physic parameters; 4) ICIS equipment check, execution of the program confirming correct connection of the temperature and neutron flux monitoring sensors; 5) monitoring of the core TPh and NPh parameters in all operating modes; 6) monitoring of the fuel condition in the core and 7) FE leak tightness monitoring at the operating reactor

  9. Calculation support for industrial production of cobalt-60 at Leningrad NPP

    International Nuclear Information System (INIS)

    Artemov, Vladimir; Elshin, Alexander; Ivanov, Alexander; Gorbunov, Evgeny; Ikonnikov, Roman; Pimenov, Alexander

    2008-01-01

    Cobalt-60 is industrially produced at the Leningrad NPP by irradiation of cobalt-59 in special-purpose facilities loaded into the RBMK reactor core (all 4 units). The paper describes calculation methods used to determine the current activity of cobalt in irradiation assemblies for their timely unloading. The described peculiarities of core calculation model account for continuous refueling, overloading of irradiation assemblies and individual thermohydraulics in each channel under variation of reactor power. Fuel burnup in the core is calculated with a time step of about 24 hours. The resulting values for cobalt activity and uncertainties are presented in the paper as well. Deviation of calculated cobalt activity from measured activity is within the experimental accuracy of 10% (at confidence probability of 0.95). (authors)

  10. Dinamics of radiating conditions in the natural complexes of a near zone of the Chernobyl NPP

    International Nuclear Information System (INIS)

    Bondar', Yu.I.; Kalinichenko, S.A.; Marchenko, Yu.D.

    2010-01-01

    The analysis of radiating conditions in exclusion zone of Chernobyl NPP is resulted. Annual and seasonal fluctuations of capacity of a dose, and also change of parameters of vertical migration Cs 137, Sr 90, Am 241 in a vertical profile of soils of various territorially-natural complexes are considered. (authors)

  11. Burnup measurements at the RECH-1 research reactor

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  13. Burnup credit feasibility for BWR spent fuel shipments

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1990-01-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  15. Determination of reactor fuel burnup using passive neutron assay

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Yan Weihua; Zhang Zhao; Xiao Zhigang; Zhang Liguo

    2014-01-01

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

  17. DRAGON and CORD-2 nuclear calculation of the NPP Krško fuel assembly

    International Nuclear Information System (INIS)

    Kromar, Marjan; Kurinčič, Bojan

    2012-01-01

    Highlights: ▶ Comparison of the DRAGON 4 and CORD-2 calculation of the NPP Krško 16 × 16 fuel assembly has been performed. ▶ Two different enrichments (4.60% and 4.95%), three IFBA configurations (0, 64 and 116 IFBAs) and the burnup up to 60,000 MWd/tU were considered. ▶ Comparison shows that the agreement in the results of both codes is very good. ▶ Regarding efficiency, one CORD-2 assembly depletion calculation takes about 3 min of the CPU time on the typical PC, while the DRAGON calculation runs more than 25 h. - Abstract: The geometry of the reactor core is usually too complex to be solved in one step. Therefore, a solution for the whole core in 3-D geometry is sought in several steps, where some kind of homogenization procedure of neutron few-group cross sections is applied. Usually, assembly-homogenized effective two-group cross sections are determined, which are suitable for solving the diffusion equation for the whole core by a coarse mesh nodal methods. In this paper DRAGON 4 and CORD-2 codes are used for the calculation of NPP Krško 16 × 16 fuel assemblies without and with IFBA rods. DRAGON code was selected, since it can use the same cross-section library as the WIMS-D5 code employed in the CORD-2 system. Different results arise therefore solely from the different models used in the calculations. The heterogeneous depletion calculation was performed up to burnup of 60,000 MWd/tU. Results of both codes are compared for the infinite multiplication factor, fast to thermal spectral ratio and pin power distributions.

  18. Fuel design and operational experience in Loviisa NPP, future trends in fuel issues

    International Nuclear Information System (INIS)

    Terasvirta, R.

    2001-01-01

    This paper summarizes the past operational experience of nuclear fuel with reference to most significant design changes during the years. In general, the fuel behaviour in Loviisa NPP in terms of leaking fuel assemblies has been good. The major improvements by fuel design changes in Lovissa NPP, including rod elongation margin, change in the pellet design and manufacturing process, upper grid modifications, change of material in the spacer grids and reduction of the shroud tube thickness are discussed and related to the number of failed fuel assemblies. The detailed investigation of fuel failure rates as function of different fuel and operation characteristics allows to classify the leaking fuel assemblies according to the cause of failure. In a brief discussion concerning new changes in the safety guide for nuclear design limits, re-issued by the Finnish Safety Authority (STUK), the frequencies for class 1 and class 2 accidents are determined. Another change in this guide is the introduction of design limits for the number of fuel rods experiencing DNB in class 1 accidents and number of failed rods in class 2 accidents. It is concluded that as far as normal operation is concerned, there seems to be sufficiently large margin between present operational limits in Loviisa and the design limits. The real limits do not come from fuel behaviour in the normal operation or operational occurrences but from the accident behaviour. At the moment, fuel assembly burnup extension beyond 45 MWd/kgU is clearly out of the question before further information and positive results are obtained on high burnup fuel behaviour in accident conditions

  19. On-line maintenance at Cofrentes NPP

    International Nuclear Information System (INIS)

    Roldan Vilches, J.; Moreno Matarranz, M. A.; Hermana Mendioroz, I.

    1998-01-01

    Cofrentes NPP has begun in 1997 activities related to At Power Preventive Maintenance over trains or systems which lead to a voluntary entry in a Limitative Condition of Operation (LCO) of the Technical Specifications. From others benefits, this program ha improved the risk management and the staff's knowledge over the functions and safety implications of the different systems, the better exploit of the resources, the co-ordination of the different organisations involved (Maintenance an Operation) and the reductions of works during shutdowns. Previous to each work, a feasibility study analyzes qualitative and quantitative (PSA), using the Risk Monitor, the implications on safety of all the tasks, assuring that the global safety of the Plant is always maintained. Tech. Spec. are analyzed in detail and also are analyzed situations of simultaneous unavailabilities of systems which could lead to a high risk situation. Two different risk controls are defined (punctual and accumulated) to assure that high risk situations will not be given. Finally, historical risk profile is analyzed to assure that the accumulated risk increase is not significant. Risk Monitor helps staff in the schedule and follow-up of the activities of On-Line Maintenance. Each one of the tasks are deeply planned and harshly analyzed and are carried out by high qualified workers. By the moment, this program is running with fully satisfaction on the Plant. (Author)

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

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

    2010-10-01

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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  2. M5TM alloy high burnup behavior and worldwide licensing

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  3. Comparison of analysis methods for burnup credit applications

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  4. Does climate directly influence NPP globally?

    Science.gov (United States)

    Chu, Chengjin; Bartlett, Megan; Wang, Youshi; He, Fangliang; Weiner, Jacob; Chave, Jérôme; Sack, Lawren

    2016-01-01

    The need for rigorous analyses of climate impacts has never been more crucial. Current textbooks state that climate directly influences ecosystem annual net primary productivity (NPP), emphasizing the urgent need to monitor the impacts of climate change. A recent paper challenged this consensus, arguing, based on an analysis of NPP for 1247 woody plant communities across global climate gradients, that temperature and precipitation have negligible direct effects on NPP and only perhaps have indirect effects by constraining total stand biomass (Mtot ) and stand age (a). The authors of that study concluded that the length of the growing season (lgs ) might have a minor influence on NPP, an effect they considered not to be directly related to climate. In this article, we describe flaws that affected that study's conclusions and present novel analyses to disentangle the effects of stand variables and climate in determining NPP. We re-analyzed the same database to partition the direct and indirect effects of climate on NPP, using three approaches: maximum-likelihood model selection, independent-effects analysis, and structural equation modeling. These new analyses showed that about half of the global variation in NPP could be explained by Mtot combined with climate variables and supported strong and direct influences of climate independently of Mtot , both for NPP and for net biomass change averaged across the known lifetime of the stands (ABC = average biomass change). We show that lgs is an important climate variable, intrinsically correlated with, and contributing to mean annual temperature and precipitation (Tann and Pann ), all important climatic drivers of NPP. Our analyses provide guidance for statistical and mechanistic analyses of climate drivers of ecosystem processes for predictive modeling and provide novel evidence supporting the strong, direct role of climate in determining vegetation productivity at the global scale. © 2015 John Wiley & Sons Ltd.

  5. NOAA JPSS Ozone Mapping and Profiler Suite (OMPS) Nadir Profile Science Sensor Data Record (SDR) from IDPS

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Ozone Mapping and Profiler Suite (OMPS) onboard the Suomi-NPP satellite monitors ozone from space. OMPS will collect total column and vertical profile ozone data...

  6. ESTE EMO and ESTE EBO - emergency response system for NPP Mochovce and NPP Bohunice V-2

    International Nuclear Information System (INIS)

    Caeny, P.; Chyly, M.; Suchon, D.; Smejkalova, E.; Fabova, V.; Mancikova, M.; Muller, P.

    2009-01-01

    Programs ESTE EMO and ESTE EBO are emergency response systems that help the crisis staff of the NPP in assessing the source term (predicted possible release of radionuclides to the atmosphere ), in assessing the urgent protective measures and sectors under threat, in assessing real release (symptoms of release really detected and observed), in calculating radiological impacts of real release, averted or avertable doses, potential doses and doses during transport or evacuation on specified routes. Both systems serve as instruments in case of severe accident (DBA or BDBA) at NPP Mochovce or NPP Bohunice, accidents with threat of release of radioactivity to the atmosphere. Systems are implemented at emergency centre of Mochovce NPP and Bohunice NPP and connected online to the sources of technological and radiological data from the reactor, primary circuit, confinement, secondary circuit, ventilation stack, from the area of NPP (TDS 1) and from the emergency planning zone (TDS 11). Systems are connected online to the sources of meteorological data, too. (authors)

  7. ESTE EMO and ESTE EBO - emergency response system for NPP Mochovce and NPP Bohunice V-2

    International Nuclear Information System (INIS)

    Caeny, P.; Chyly, M.; Suchon, D.; Smejkalova, E.; Fabova, V.; Mancikova, M.; Muller, P.

    2008-01-01

    Programs ESTE EMO and ESTE EBO are emergency response systems that help the crisis staff of the NPP in assessing the source term (predicted possible release of radionuclides to the atmosphere ), in assessing the urgent protective measures and sectors under threat, in assessing real release (symptoms of release really detected and observed), in calculating radiological impacts of real release, averted or avertable doses, potential doses and doses during transport or evacuation on specified routes. Both systems serve as instruments in case of severe accident (DBA or BDBA) at NPP Mochovce or NPP Bohunice, accidents with threat of release of radioactivity to the atmosphere. Systems are implemented at emergency centre of Mochovce NPP and Bohunice NPP and connected online to the sources of technological and radiological data from the reactor, primary circuit, confinement, secondary circuit, ventilation stack, from the area of NPP (TDS 1) and from the emergency planning zone (TDS 11). Systems are connected online to the sources of meteorological data, too. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-01

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

  9. Study on human factor at NPP

    International Nuclear Information System (INIS)

    Nopp, I.

    1984-01-01

    Factors affecting the reliabilty of the reactor control by an NPP operator are considered on the base of the Czechoslovakia NPP operating experience. The reliability level of NPP operators depends on objective factors (conditions and regime of labour) determining the labour productivity and on subjective ones (psychological morale, physical and mental abilities and occupational level of personnel). Problems of the effect of physical and mental abilities and professional level on the reliability of personnel are considered to be the most important ones. The effect of individual abilities and specific features of the human body on changes in his occupational abilities can be estimated only to a certain degree

  10. Inspection Qualification Centre in NPP 'Kozloduy'

    International Nuclear Information System (INIS)

    Mikhovski, M.

    2000-01-01

    In May 1999 according to the working plan of the IAEA project RER 4/020 and the decision of the NPP the Inspection Qualification Centre (IQC) has been established in order to provide examination services in the NPP. During year 1999 IVC (AEA Technology) in the framework of the DTI project provides consulting and technical assistance to the NPP, IQC, Bulgarian Academy of Sciences and Regulatory Authorities in setting up the IQC infrastructure. Now IQC work as an independent inspection body B type. The IQC activities for the period 1999-2000 are presented and further tasks are outlined

  11. A brief overview of Ignalina NPP safety issues

    International Nuclear Information System (INIS)

    Almenas, K.; Ushpuras, E.

    1998-01-01

    A description of the safety of Ignalina NPP in a very popular form is presented. Answers to the most frequently recurring questions concerning the Ignalina NPP are provided based on recently completed international studies. Questions are like these: can a similar accident to the one that occurred in Chernobyl take place at Ignalina NPP, does the Ignalina NPP have a containment, what are the probabilities and potential consequences of accidents, etc. The brochure contains a short description of Ignalina NPP safety improvement programs

  12. 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)

  13. Krsko NPP radioactive waste characteristics

    International Nuclear Information System (INIS)

    Skanata, D.; Kroselj, V.; Jankovic, M.

    2007-01-01

    In May 2005 Krsko NPP initiated the Radioactive Waste Characterization Project and commissioned its realization to the consulting company Enconet International, Zagreb. The Agency for Radwaste Management was invited to participate on the Project. The Project was successfully closed out in August 2006. The main Project goal consisted of systematization the existing and gathering the missing radiological, chemical, physical, mechanical, thermal and biological information and data on radioactive waste. In a general perspective, the Project may also be considered as a part of broader scope of activities to support state efforts to find a disposal solution for radioactive waste in Slovenia. The operational low and intermediate level radioactive waste has been structured into 6 waste streams that contain evaporator concentrates and tank sludges, spent ion resins, spent filters, compressible and non-compressible waste as well as specific waste. For each of mentioned waste streams, process schemes have been developed including raw waste, treatment and conditioning technologies, waste forms, containers and waste packages. In the paper the main results of the Characterization Project will be briefly described. The results will indicate that there are 17 different types of raw waste that have been processed by applying 9 treatment/conditioning technologies. By this way 18 different waste forms have been produced and stored into 3 types of containers. Within each type of container several combinations should be distinguished. Considering all of this, there are 34 different types of waste packages altogether that are currently stored in the Solid Radwaste Storage Facility at the Krsko NPP site. Because of these findings a new identification system has been recommended and consequently the improvement of the existing database on radioactive waste has been proposed. The potential areas of further in depth characterization are indicated. In the paper a brief description on the

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  16. New generation of CASTOR registered casks for high enriched, high burn-up fuel from German NPP

    International Nuclear Information System (INIS)

    Gartz, R.; Kuehne, B.; Diersch, R.

    2004-01-01

    Requirements for new cask designs for transport and long-term dry storage of spent fuel assemblies (FA) from LWR-reactors are based on both increased source terms of the LWR FA including MOX FA, as well as the demand for economical optimisation of decommissioning costs by increased cask capacities. For this, cask development is the challenge to create and establish cask designs that can accommodate more FA with higher source terms, each under fixed boundary conditions (i.e. transport requirements and limitations of the power plants as crane loads and/or fixed maximum dimensions). This task has been elaborated by working simultaneously on different development actions each focussed to improve the cask performance. In the following a brief summary will be presented to give an overview which developments and investigations have been and are still will be performed for development and safety analyses of the new CASTOR registered -designs under the main subjects: material investigation and qualification, component tests and verifications, detailed design analysis and not at least design verification

  17. Power ramp tests of high burnup BWR segment rods

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. Treatment of NPP wastes using vitrification

    International Nuclear Information System (INIS)

    Sobolev, I.A.; Lifanov, F.A.; Stefanovsky, S.V.; Kobelev, A.P.; Savkin, A.E.; Kornev, V.I.

    1998-01-01

    Glass-based materials to immobilize various liquid and solid radioactive wastes generated at nuclear power plants (NPP) were designed. Glassy waste forms can be produced using electric melting including a cold crucible melting. Leach rate of cesium was found to be 10 -5 -10 -6 g/(cm 2 day) (IAEA technique). Volume reduction factor after vitrification reached 4-5. Various technologies for NPP waste vitrification were developed. Direct vitrification means feeding of source waste into the melter with formation of glassy waste form to be disposed. Joule heated ceramic melter, and cold crucible were tested. Process variables at treatment of Kursk, Chernobyl (RBMK), Kalinin, Novovoronezh (VVER) NPP wastes were determined. The most promising melter was found to be the cold crucible. Pilot plant based on the cold crucibles has been designed and constructed. Solid burnable NPP wastes are incinerated and slags are incorporated in glass. (author)

  19. Environmental impact assessment of NPP decommissioning

    International Nuclear Information System (INIS)

    Hinca, R.

    2009-01-01

    In this presentation the following potential impacts of decommissioning of NPP are discussed: - Impacts on population; Impacts on natural environment; Land impacts; Impacts on urban complex and land utilisation; Possible impacts on area as a result of failure.

  20. Radiation ecological monitoring in NPP region

    International Nuclear Information System (INIS)

    Egorov, Yu.A.; Kazakov, S.V.

    1985-01-01

    The known principle of sanitary-hygienic regulation of NPP radiation effect on man and environment is analyzed. An ecological approach is required to optimize NPP relations with the environment and to regulate radioactivity of the NPP - environment system. The ecological approach envisages the development of standards of permissible concentrations of radioactive and chemical substances (as well as heat) in natural environment, taking into account their synergism, corresponding to ecologically permissible response reactions of biota to their effect. The ecological approach also comprises the sanitary-hygienic principle of radiation protection of man. Attention is paid to ecological monitoring in NPP region, comprising consideration of factors, affecting the environment, evaluation of the actual state of the environment, prediction of the environmental state, evaluation of the expected environmental state

  1. NPP operation and modern high technologies

    International Nuclear Information System (INIS)

    Kuznetsov, V.V.

    1992-01-01

    Examples are considered of modern high technology introduction into daily practice of NPP operation, namely: satellite communication systems, robots, non-destructive testing, optical-fiber techniques, laser measuring means and others

  2. Krsko NPP Periodic Safety Review program

    International Nuclear Information System (INIS)

    Basic, I.; Spiler, J.; Novsak, M.

    2001-01-01

    The need for conducting a Periodic Safety Review for the Krsko NPP has been clearly recognized both by the NEK and the regulator (SNSA). The PSR would be highly desirable both in the light of current trends in safety oversight practices and because of many benefits it is capable to provide. On January 11, 2001 the SNSA issued a decision requesting the Krsko NPP to prepare a program and determine a schedule for the implementation of the program for 'Periodic Safety Review of NPP Krsko'. The program, which is required to be in accordance with the IAEA safety philosophy and with the EU practice, was submitted for the approval to the SNSA by the end of March 2001. The paper summarizes Krsko NPP Periodic Safety Review Program [1] including implemented SNSA and IAEA Expert Mission comments.(author)

  3. NPP A-1 decommissioning - Phase I

    International Nuclear Information System (INIS)

    Krstenik, A.; Blazek, J.

    2000-01-01

    Nuclear power plant A-1 with output 150 MW e , with metallic natural uranium fuelled, CO 2 cooled and heavy water moderated reactor had been prematurely finally shut down in 1977. It is necessary to mention that neither operator nor regulatory and other authorities have been prepared for the solution of such situation. During next two consecutive years after shutdown main effort of operator focused on technical and administrative activities which are described in the previous paper together with approach, condition and constraints for NPP A-1 decommissioning as well as the work and research carried out up to the development and approval of the Project for NPP A-1 decommissioning - I. phase. Subject of this paper is description of: (1) An approach to NPP A -1 decommissioning; (2) An approach to development of the project for NPP A-1 decommissioning; (3) Project - tasks, scope, objectives; (4) Mode of the Project realisation; (5) Progress achieved up to the 1999 year. (authors)

  4. Nuclear Power Plant (NPP) safety in Brazil

    International Nuclear Information System (INIS)

    Lederman, L.

    1980-01-01

    The multidisciplinary aspects of the activities involved in the nuclear power plant (NPP) licensing, are presented. The activities of CNEN's technical staff in the licensing of Angra-1 and Angra-2 power plants are shown. (E.G.) [pt

  5. Cuba: Juragua NPP. Project Control. Annex 5

    International Nuclear Information System (INIS)

    Serradet, M.A.

    1999-01-01

    This annex deals with project control. The long suspension of Juragua NPP has affected personnel, equipment and site structures. Efforts are being made to revive the plant and to protect existing resources (assets). An action plan has been prepared. (author)

  6. Trends and results in In-Core management for the Kozloduy NPP WWER-440 reactors

    International Nuclear Information System (INIS)

    Haralampieva, Tz.; Antov, A.; Georgieva, N.; Spasova, V.

    2001-01-01

    The paper presents the experience gained during the design and operation of the last fuel cycles of the four WWER-440/V-230 units at Kozloduy NPP. High efficiency and economy of the fuel utilization requires very precise procedures for fuel in-core management, including calculations and analyses for reloading scheme design, compared with results from operational measurements and fuel cycle efficiency. The paper describes the main stages of implementation of advanced fuel assemblies in the Kozloduy NPP WWER-440 reactors. New advanced fuel has been implemented after the completion of comprehensive neutron-physical, thermal-hydraulic and thermal-mechanical analyses by using advanced computer codes. As a general task of the fuel cycle improvements it is pointed the increasing of the final fuel burnup and decreasing of the number of spent fuel assemblies. Series of calculations and analyses, related to the introducing of the advanced fuel assemblies and improvement of the fuel cycle characteristics have been carried out to guarantee the safe operation and fuel reliability

  7. Training human resource for NPP in Vietnam

    International Nuclear Information System (INIS)

    Nguyen, Trung Tinh; Dam, Xuan Hiep

    2008-01-01

    Vietnam will establish the first NPP in the near future. With us the first important thing is the human resource, but now there is no university in Vietnam training nuclear engineers. In EPU (Electric Power University), now we are preparing for training nuclear engineers. In this paper, we review the nuclear man power and the way to train the high quality human resource for NPP and for other nuclear application in Vietnam. (author)

  8. Ignalina NPP Safety Analysis: Models and Results

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

    Research directions, linked to safety assessment of the Ignalina NPP, of the scientific safety analysis group are presented: Thermal-hydraulic analysis of accidents and operational transients; Thermal-hydraulic assessment of Ignalina NPP Accident Localization System and other compartments; Structural analysis of plant components, piping and other parts of Main Circulation Circuit; Assessment of RBMK-1500 reactor core and other. Models and main works carried out last year are described. (author)

  9. Heat delivery from Bohunice NPP, Slovakia

    International Nuclear Information System (INIS)

    Paley, I.

    1998-01-01

    Experience with nuclear district heating in the Slovak Republic is reported. The heating system of the town of Trnava is supplied by the Bohunice NPP and conventional sources. Construction of the hot water heating system from the Bohunice NPP began in 1983. Commercial operation began on 10 December 1987. Heat delivery has gradually increased from 478 TJ in 1988, to 1,104 TJ in 1995. The heat cost is low, resulting in an increasing number of consumers. (author)

  10. Radionuclide localization at the Chernobyl' NPP territory

    International Nuclear Information System (INIS)

    Mamaev, L.A.; Galkin, G.A.; Khrabrov, S.L.; Polyakov, A.S.; Mikhejkin, S.V.

    1989-01-01

    Experience is generalized of using different dust suppression (DS) compounds during Chernobyl' accident consequence elimination. Polymer DS compounds were used at the NPP operating site; the compounds kept dust-like radioactive contaminations during 1-2 months. DS at the country was realized by means of the compound on base of latex. The conclusion is made that the DS measures improved radiation situation in the NPP zone. 7 refs

  11. Knowledge Management and Organizational Proficiency with NPP

    International Nuclear Information System (INIS)

    Marler, M.

    2016-01-01

    Full text: The pace of new NPP construction, startup, and operation is straining the supply of proficient operators, technicians, and engineers. This technical brief explains an approach implemented by a US nuclear utility to capture and transfer knowledge possessed by proficient workers to new workers using the VISION learning content management system. This approach could also be used to accelerate worker proficiency in new NPP organizations. (author

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

    International Nuclear Information System (INIS)

    Bartosek, V.; Zalesky, K.

    1976-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Hesse, Ulrich; Sieberer, Johann

    2006-01-01

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

  15. Nuclide Importance and the Steady-State Burnup Equation

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Nemoto, Atsushi

    2000-01-01

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

  16. Suomi NPP Ground System Performance

    Science.gov (United States)

    Grant, K. D.; Bergeron, C.

    2013-12-01

    The National Oceanic and Atmospheric Administration (NOAA) and National Aeronautics and Space Administration (NASA) are jointly acquiring the next-generation civilian weather and environmental satellite system: the Joint Polar Satellite System (JPSS). JPSS will replace the afternoon orbit component and ground processing system of the current Polar-orbiting Operational Environmental Satellites (POES) managed by NOAA. The JPSS satellites will carry a suite of sensors designed to collect meteorological, oceanographic, climatological and geophysical observations of the Earth. The first satellite in the JPSS constellation, known as the Suomi National Polar-orbiting Partnership (Suomi NPP) satellite, was launched on 28 October 2011, and is currently undergoing product calibration and validation activities. As products reach a beta level of maturity, they are made available to the community through NOAA's Comprehensive Large Array-data Stewardship System (CLASS). CGS's data processing capability processes the satellite data from the Joint Polar Satellite System satellites to provide environmental data products (including Sensor Data Records (SDRs) and Environmental Data Records (EDRs)) to NOAA and Department of Defense (DoD) processing centers operated by the United States government. CGS is currently processing and delivering SDRs and EDRs for Suomi NPP and will continue through the lifetime of the Joint Polar Satellite System programs. Following the launch and sensor activation phase of the Suomi NPP mission, full volume data traffic is now flowing from the satellite through CGS's C3, data processing, and data delivery systems. Ground system performance is critical for this operational system. As part of early system checkout, Raytheon measured all aspects of data acquisition, routing, processing, and delivery to ensure operational performance requirements are met, and will continue to be met throughout the mission. Raytheon developed a tool to measure, categorize, and

  17. CARA design criteria for HWR fuel burnup extension

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. Effect of fuel burnup on the mechanical safety coefficients

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  19. Validation of SCALE-4 for burnup credit applications

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-09-11

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

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

    International Nuclear Information System (INIS)

    White, J.R.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Anderson, D.; Battistoni, P.

    1987-01-01

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

  3. Investigation of Burnup Credit Issues in BWR Fuel

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Haliyati R, Evi; Su'ud, Zaki

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Seifritz, W.

    2007-01-01

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

  7. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    1981-03-01

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

  9. Operation Aspect of the Main Control Room of NPP

    International Nuclear Information System (INIS)

    Sahala M Lumbanraja

    2009-01-01

    The main control room of Nuclear Power Plant (NPP) is operational centre to control all of the operation activity of NPP. NPP must be operated carefully and safely. Many aspect that contributed to operation of NPP, such as man power whose operated, technology type used, ergonomic of main control room, operational management, etc. The disturbances of communication in control room must be anticipated so the high availability of NPP can be achieved. The ergonomic of the NPP control room that will be used in Indonesia must be designed suitable to anthropometric of Indonesia society. (author)

  10. Decommissioning Study of Oskarshamn NPP

    International Nuclear Information System (INIS)

    Larsson, Helena; Anunti, Aake; Edelborg, Mathias

    2013-06-01

    By Swedish law it is the obligation of the nuclear power utilities to satisfactorily demonstrate how a nuclear power plant can be safely decommissioned and dismantled when it is no longer in service as well as calculate the estimated cost of decommissioning of the nuclear power plant. Svensk Kaernbraenslehantering AB (SKB) has been commissioned by the Swedish nuclear power utilities to meet the requirements of current legislation by studying and reporting on suitable technologies and by estimating the costs of decommissioning and dismantling of the Swedish nuclear power plants. The present report is an overview, containing the necessary information to meet the above needs, for Oskarshamn NPP. Information is given for the plant about the inventory of materials and radioactivity at the time for final shutdown. A feasible technique for dismantling is presented and the waste management is described and the resulting waste quantities are estimated. Finally a schedule for the decommissioning phase is given and the costs associated are estimated as a basis for funding

  11. Decommissioning study of Forsmark NPP

    International Nuclear Information System (INIS)

    Anunti, Aake; Larsson, Helena; Edelborg, Mathias

    2013-06-01

    By Swedish law it is the obligation of the nuclear power utilities to satisfactorily demonstrate how a nuclear power plant can be safely decommissioned and dismantled when it is no longer in service as well as calculate the estimated cost of decommissioning of the nuclear power plant. Svensk Kaernbraenslehantering AB (SKB) has been commissioned by the Swedish nuclear power utilities to meet the requirements of current legislation by studying and reporting on suitable technologies and by estimating the costs of decommissioning and dismantling of the Swedish nuclear power plants. The present report is an overview, containing the necessary information to meet the above needs, for the Forsmark NPP. Information is given for the plant about the inventory of materials and radioactivity at the time for final shutdown. A feasible technique for dismantling is presented and the waste management is described and the resulting waste quantities are estimated. Finally a schedule for the decommissioning phase is given and the costs associated are estimated as a basis for funding

  12. ORSEC technological risks, Blayais NPP

    International Nuclear Information System (INIS)

    2016-01-01

    The Particular intervention plan (PPI in French) is an emergency plan which foresees the measures and means to be implemented to address the potential risks of the presence and operation of a nuclear facility. This plan is implemented and developed by the Prefect in case of nuclear accident (or incident leading to a potential accident), the impact of which extending beyond the facility perimeter. It represents a special section of the organisation plan for civil protection response (ORSEC plan). The PPI foresees the necessary measures and means for crisis management during the first hours following the accident and is triggered by the Department Prefect according to the information provided by the facility operator. Its aim is to protect the populations leaving within 10 km of the facility against a potential radiological hazard. The PPI describes: the facility, the intervention area, the protection measures for the population, the conditions of emergency plan triggering, the crisis organisation, the action forms of the different services, and the post-accident stage. This document is the public version of the Particular intervention plan of the Blayais NPP (Gironde, France)

  13. Decommissioning Study of Oskarshamn NPP

    Energy Technology Data Exchange (ETDEWEB)

    Larsson, Helena; Anunti, Aake; Edelborg, Mathias [Westinghouse Electric Sweden AB, Vaesteraas (Sweden)

    2013-06-15

    By Swedish law it is the obligation of the nuclear power utilities to satisfactorily demonstrate how a nuclear power plant can be safely decommissioned and dismantled when it is no longer in service as well as calculate the estimated cost of decommissioning of the nuclear power plant. Svensk Kaernbraenslehantering AB (SKB) has been commissioned by the Swedish nuclear power utilities to meet the requirements of current legislation by studying and reporting on suitable technologies and by estimating the costs of decommissioning and dismantling of the Swedish nuclear power plants. The present report is an overview, containing the necessary information to meet the above needs, for Oskarshamn NPP. Information is given for the plant about the inventory of materials and radioactivity at the time for final shutdown. A feasible technique for dismantling is presented and the waste management is described and the resulting waste quantities are estimated. Finally a schedule for the decommissioning phase is given and the costs associated are estimated as a basis for funding.

  14. Decommissioning study of Forsmark NPP

    Energy Technology Data Exchange (ETDEWEB)

    Anunti, Aake; Larsson, Helena; Edelborg, Mathias [Westinghouse Electric Sweden AB, Vaesteraas (Sweden)

    2013-06-15

    By Swedish law it is the obligation of the nuclear power utilities to satisfactorily demonstrate how a nuclear power plant can be safely decommissioned and dismantled when it is no longer in service as well as calculate the estimated cost of decommissioning of the nuclear power plant. Svensk Kaernbraenslehantering AB (SKB) has been commissioned by the Swedish nuclear power utilities to meet the requirements of current legislation by studying and reporting on suitable technologies and by estimating the costs of decommissioning and dismantling of the Swedish nuclear power plants. The present report is an overview, containing the necessary information to meet the above needs, for the Forsmark NPP. Information is given for the plant about the inventory of materials and radioactivity at the time for final shutdown. A feasible technique for dismantling is presented and the waste management is described and the resulting waste quantities are estimated. Finally a schedule for the decommissioning phase is given and the costs associated are estimated as a basis for funding.

  15. 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

  16. Changing NPP consumption patterns in the Holocene: from Megafauna "liberated" NPP to "ecological bankruptcy"

    Science.gov (United States)

    Doughty, C.

    2015-12-01

    There have been vast changes in how net primary production (NPP) is consumed by humans and animals during the Holocene beginning with a potential increase in availability following the Pleistocene megafauna extinctions. This was followed by the development of agriculture which began to gradually restrict availability of NPP for wild animals. Finally, humans entered the industrial era using non-plant based energies to power societies. Here I ask the following questions about these three energy transitions: 1. How much NPP energy may have become available following the megafauna extinctions? 2. When did humans, through agriculture and domestic animals, consume more NPP than wild mammals in each country? 3. When did humans and wild mammals use more energy than was available in total NPP in each country? To answer this last question I calculate NPP consumed by wild animals, crops, livestock, and energy use (all converted to units of MJ) and compare this with the total potential NPP (also in MJ) for each country. We develop the term "ecological bankruptcy" to refer to the level of consumption where not all energy needs can be met by the country's NPP. Currently, 82 countries and a net population of 5.4 billion are in the state of ecologically bankruptcy, crossing this threshold at various times over the past 40 years. By contrast, only 52 countries with a net population of 1.2 billion remain ecologically solvent. Overall, the Holocene has seen remarkable changes in consumption patterns of NPP, passing through three distinct phases. Humans began in a world where there was 1.6-4.1% unclaimed NPP to consume. From 1700-1850, humans began to consume more than wild animals (globally averaged). At present, >82% of people live in countries where not even all available plant matter could satisfy our energy demands.

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

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

    2002-02-01

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

  18. 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)

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Zagar, Tomaz; Ravnik, Matjaz

    2000-01-01

    This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

  1. Automatic control system at the ''Loviisa'' NPP

    International Nuclear Information System (INIS)

    Kukhtevich, I.V.; Mal'tsev, B.K.; Sergievskaya, E.N.

    1980-01-01

    Automatic control system of the Loviisa-1 NPP (Finland) is described. According to operation conditions of Finland power system the Loviisa-1 NPP must operate in the mode of week and day control of loading schedule and participate in current control of power system frequency and capacity. With provision for these requirements NPP is equipped with the all-regime system for automatic control functioning during reactor start-up, shut-down, in normal and transient regimes and in emergency situations. The automatic control system includes: a data subsystem, an automatic control subsystem, a discrete control subsystem including remote, a subsystem for reactor control and protection and overall station system of protections: control and dosimetry inside the reactor. Structures of a data-computer complex, discrete control subsystems, reactor control and protection systems, neutron flux control system, inside-reactor control system, station protection system and system for control of fuel element tightness are presented in short. Two-year experience of the NPP operation confirmed advisability of the chosen volume of automatization. The Loviisa-1 NPP operates successfully in the mode of the week and day control of supervisor schedule and current control of frequency (short-term control)

  2. Environmental impact of the NPP Krsko

    International Nuclear Information System (INIS)

    Novosel, N.

    1996-01-01

    The Ministry of Economic Affairs has for six years now been monitoring the operation of the Krsko NPP (NEK) and its impact on the environment. A bulletin titled 'NEK - Energy and Environment' is being issued every three months. It contains information on operation of the Krsko NPP for the previous three months, a graph of duration of temperature increase of water in the Sava river (delta T) in that period, an assessment of the radiological impact of Krsko NPP on the environment through an equivalent dose cumulatively throughout the calendar year, and a short current text related to Krsko NPP. The Ministry of Economic Affairs organizes a press conference on every issue of the bulletin, as an attempt of introducing this subject to the media and to the public. This paper contains a review of information given in the NEK bulletin from 1990 to 1995 with a special emphasis on the contribution of the Krsko NPP to the artificially caused radiation on the border between the Republic of croatia and the Republic of Slovenia. (author)

  3. Integrated ageing management of Atucha NPP

    International Nuclear Information System (INIS)

    Ranalli, Juan M.; Marchena, Martin H.; Zorrilla, Jorge R.; Antonaccio, Elvio E.; Brenna, Pablo; Yllanez, Daniela; Cruz, Gerardo Vera de la; Luraschi, Carlos; Sabransky, Mario

    2013-01-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor, twice as big as Atucha I finishing a delayed construction. With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  4. Integrated Ageing Management of Atucha NPP

    International Nuclear Information System (INIS)

    Ranalli, J.M.; Marchena, M.H.; Zorrilla, J.R.; Sabransky, M.

    2012-01-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor twice as big as Atucha I finishing a delayed construction . With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  5. Integrated ageing management of Atucha NPP

    Energy Technology Data Exchange (ETDEWEB)

    Ranalli, Juan M.; Marchena, Martin H.; Zorrilla, Jorge R.; Antonaccio, Elvio E.; Brenna, Pablo; Yllanez, Daniela; Cruz, Gerardo Vera de la; Luraschi, Carlos, E-mail: ranalli@cnea.gov.ar [Gerencia Coordinacion Proyectos CNEA-NASA, Comision Nacional de Energia Atomica, Buenos Aires (Argentina); Sabransky, Mario, E-mail: msabransky@na-sa.com.ar [Departamento Gestion de Envejecimiento, Central Nuclear Atucha I-II Nucleoelectrica Argentina S.A., Provincia de Buenos Aires (Argentina)

    2013-07-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor, twice as big as Atucha I finishing a delayed construction. With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  6. Polynomial expansion methodology for microscopic cross sections to use in spatial burnup calculations

    International Nuclear Information System (INIS)

    Conti Filho, P.; Oliveira Barroso, A.C. de

    1985-01-01

    It was developed a computer code to generate polynomial coefficients which represent homogenized microscopic cross sections in function of the local accumulated burnup and concentration of soluble boron, presented in fuel element, for each step of burnup reactor. Afterward, it was developed a coupling between LEOPARD-GERADOR DE POLINOMIOS - CITATION computer codes to interpret and build homogenized microscopic cross sections according with local characteristics of each fuel element during the burnup calculation of reactor core. (M.C.K.) [pt

  7. Comparison of measured and calculated burn-up of AVR-Fuel-Elements

    Energy Technology Data Exchange (ETDEWEB)

    Wagemann, R.

    1974-03-15

    Burn-up comparisons are made for small batches of three types of AVR fuel elements using a coupled EREBUS-MUPO neutronic analysis compared against test results from both nondestructive gamma-ray measurements of cesium-137 activity and destructive mass spectrometry measurements of the ratio of U-233 to U-235. The comparisons are relatively good for average burn-up and reasonably good for burn-up distributions.

  8. Effect of local burn-up variation on computed mean nuclide concentrations

    International Nuclear Information System (INIS)

    Moeller, W.

    1982-01-01

    Mean concentrations of U-235, U-236, U-238, Pu-239, Pu-240, Pu-241 and Pu-242 in some volume areas of WWER-440 fuel assemblies have been calculated from corresponding burn-up microdistribution data and compared with those calculated from burn-up mean values. Differences occurring were below 3% for the uranium nuclides but, at low burn-ups, considerable for Pu-241 and Pu-242. (author)

  9. Reserves of labour content reduction in NPP construction

    International Nuclear Information System (INIS)

    Bekerman, R.E.

    1986-01-01

    Specific labour contents when constructing NPP with RBMK-1000 and WWER-1000 type reactors are presented. Factors affecting labour content of NPP construction are shown. Measures aimed at labour content decrease are suggested

  10. Regulatory aspects of NPP safety

    International Nuclear Information System (INIS)

    Stuller, J.; Brandejs, P.; Miasnikov, A.; Svab, M.

    1999-01-01

    In beginning, a history of legislative process regulating industrial utilisation of nuclear energy is given, including detailed list of decrees issued by the first regulatory body supervising Czech nuclear installations - Czechoslovak Atomic Energy Commission (CSKAE). Current status of nuclear regulations and radiation protection, especially in connection with Atomic Act (Act No 18/1997 Coll.), is described. The Atomic Act transfers into the Czech legal system a number of obligations following from the Vienna Convention on Civil Liability for Nuclear Damage and Joint Protocol relating to the Application of the Vienna and Paris Convention, to which the Czech Republic had acceded. Actual duties and competence of current nuclear regulatory body - State Office for Nuclear Safety (SUJB) - are given in detail. Execution of the State supervision of peaceful utilisation of nuclear energy and ionising radiation is laid out in several articles of the Act, which comprises: control activities of the SUJB, remedial measures, penalties. Material and human resources are sufficient for fulfilment of the basic functions for which SUJB is authorised by the law. For 1998, the SUJB allotted staff of 149, approximately 2/3 of that number are nuclear safety and radiation protection inspectors. The SUJB budget for 1998 is approximately 180 million Czech crowns (roughly 6 million US dollars). Inspection activity of SUJB is carried out in three different ways: routine inspections, planned specialised inspections, inspections as a response to a certain situation (ad-hoc inspections). Approach to the licensing of major plant upgrades and backfittings are mainly illustrated on the Temelin NPP licensing. Regulatory position and practices concerning review activities are presented. (author)

  11. Calculation of effect of burnup history on spent fuel reactivity based on CASMO5

    International Nuclear Information System (INIS)

    Li Xiaobo; Xia Zhaodong; Zhu Qingfu

    2015-01-01

    Based on the burnup credit of actinides + fission products (APU-2) which are usually considered in spent fuel package, the effect of power density and operating history on k_∞ was studied. All the burnup calculations are based on the two-dimensional fuel assembly burnup program CASMO5. The results show that taking the core average power density of specified power plus a bounding margin of 0.0023 to k_∞, and taking the operating history of specified power without shutdown during cycle and between cycles plus a bounding margin of 0.0045 to k_∞ can meet the bounding principle of burnup credit. (authors)

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

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1986-03-01

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

  13. Improvement of coal focus and cooling towers of COFRENTES NPP

    International Nuclear Information System (INIS)

    Martinez, I.; Bogh, P.

    1998-01-01

    Cofrentes Nuclear Power Plant is performing a improving program of its cooling towers based on the filling revamping and cooling water circuit improvement. Furthermore, and as consequence of the acquired experience on cooling towers due to the mentioned program, Cofrentes NPP has decided to follow up with this project from a different point of view based on the thermal-hydraulic optimization of the cooling process inside the towers. This program, which is going to be carried out by Cofrentes NPP, Iberinco and Energy Planning and Power Generation (EPPG) provides an improvement on the thermal profile and of the draught inside the cooling towers by improving the water distribution in the towers active area. In order to perform such a program is needed to fulfill a test program to assure a guaranteed performance gain. In this way, Iberinco is developing a test procedure which improves the results which are obtained with the present standards used commonly by the industry. As a consequence of this program, Cofrentes is expecting to obtain a gain of 5 to 8 MWe with a revenue period of 4 to 5 months, results validated in another foreigner Plant which have developed a similar program. (Author)

  14. Croatian Capacity for Management of the NPP Krsko Radioactive Waste

    International Nuclear Information System (INIS)

    Levanat, Ivica; Lokner, Vladimir

    2014-01-01

    Together with Slovenia, Croatia is responsible for management of the NPP Krsko radioactive waste and spent fuel. So far, no firm agreement on specific solutions has been reached between the two countries. On the contrary, all activities related to revision and development of the joint Program of the NPP decommissioning and spent fuel and radioactive waste management were discontinued several years ago. Unless Slovenia and Croatia definitely agree on joint solutions in the meantime, Croatia will have to begin transfer of one half of the NPP Krsko spent fuel and radioactive waste into its territory in about nine years. Presently, however, Croatian capacities for such an undertaking are seriously inadequate in several respects, and they are not developing in any promising way. For no rational reason, this state of the Croatian capacities has been maintained (recently even deteriorating in some respects) for at least a decade, disregarding also for at least two years the explicit requirements of the EU Directive 2011/70/EURATOM aimed at establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste. In fall of 2012, Fund for financing the NPP Krsko decommissioning and spent fuel and radioactive waste disposal was appointed as the Croatian expert organization for revision and development of the program of these same activities it had originally been supposed only to finance. In 2013 the Fund expanded its activities to a low-profile attempt at revitalization of the Croatian radioactive waste repository project, although the Fund is not yet properly capacitated for either of these tasks. The above is hardly in compliance with the Directive requirements, such as to establish 'a national legislative, regulatory and organisational framework' 'that allocates responsibility... between relevant competent bodies'. The lack of competent experts in the field appears to affect the quality of Croatian legislative

  15. Triton burnup study using scintillating fiber detector on JT-60U

    International Nuclear Information System (INIS)

    Harano, Hideki

    1997-09-01

    The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV α particles behave. The α 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 p -scan and n e -scan experiments have been performed. The R p -scan experiment indicates that the triton's transport was increased as the ripple amplitude over the triton became larger. In the n 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

  16. 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.

  17. Research on the NPP human factors engineering operating experience review

    International Nuclear Information System (INIS)

    Ren Xiangchen; Miao Hongxing; Ning Zhonghe

    2006-01-01

    This paper addresses the importance of the human factors engineering (HFE) for the design of nuclear power plant (NPP), especially for the design of human-machine interface in the NPP. It also summarizes the scope and content of the NPP HFE. The function, scope, content and process of the NPP human factors engineering operating experience review (OER) are mainly focused on, and significantly discussed. Finally, it briefly introduces the situation of the studies on the OER in China. (authors)

  18. Some problems of NPP construction base improvement

    International Nuclear Information System (INIS)

    Movchan, S.V.

    1984-01-01

    NPP construction bases are characterized by high cost of construction and large area. Duration of base construction makes up 3-4 years, labour contents for their erection constitute 600-900 thousand man-days. Delays in organizing functional base services essentially decelerate construction rates of the main NPP buildings. Maximum joining of separate buildings by their functional assignment and structural peculiarities, wide application of container buildings, partial utilization of permanent buildings of production centre for construction needs; transition to new organizational form of construction based on industrial production of buildings; production of volumetric structural-technological cells with mounted equipment manufactured at specialized plants, mounting NPP components with stock produced cells, consideration of the problem of large power centre creation are necessary for reduction of construction centres, area reduction of cost and duration of their construction

  19. New trends in designing NPP control boards

    International Nuclear Information System (INIS)

    Kondrat'ev, V.V.

    1981-01-01

    A short analytical summary of the latest developments and future trends in designing NPP control boards is given. The designs of the Westinghause and the Hynkley-Point NPP control boards are described in detail. The essence of the advanced control board concept consists , firstly, in expanded use of computer-controlled displays for the sake of reducing the content of unimportant information presented to an operator, and, secondary, in better account of human possibilities to convert the NPP operation information into a more suitable form. An enlarged use of the direct digital reactor control utilizing microprocessors is expected. Besides, the employment of full-scale control board mock-ups and information desks as well as testing newly-developed control boards at computer reactor simulators are concluded to be used at all-growing rate [ru

  20. Qinshan NPP large break LOCA safety analysis

    International Nuclear Information System (INIS)

    Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

    1997-01-01

    Qinshan NPP is the first nuclear power plant in the mainland of China, a 300 MW(e) two-loop PWR. Large break LOCA is the design-basis accident of Qinshan NPP. Based on available computer codes, the own analysis method which complies with Appendix k of 10 CFR 50 has been established. The RELAP4/MOD7 code is employed for the calculations of blowdown, refill and reflood phase of the RCS respectively. The CONTEMPT-LT/028 code is used for the containment pressure and temperature analysis. The temperature transient in the hot rod is calculated using the FRAP-6T code. Conservative initial and functional assumptions were adopted during Qinshan NPP large break LOCA analysis. The results of the analysis show the applicable acceptance criteria for the loss-of-coolant accident are met

  1. Safeguards at Kozloduy NPP - Experience and expectations

    International Nuclear Information System (INIS)

    Elenkov, Todor

    2001-01-01

    Bulgaria is a party of Non Proliferation Treaty since 5 September 1969. The agreement between IAEA and Bulgaria - INFCIRC 178 - has been in force since 29 February 1972. At that time Bulgaria had one research reactor IRT-2000 in Sofia and two power reactors of WWER-440 type under construction. Now at Kozloduy NPP site there are 4 facilities, which consist of 4 WWER-440 and 2 WWER-1000 type power reactors, producing almost 50% of the electricity in Bulgaria and 1 wet away from reactor spent fuel storage. In 1991 under the green movements and social pressure, the research reactor in Sofia was closed and the construction of the second NPP in Belene with 2 WWER-1000 type reactors was halted. After the transfer in 1994 of the fresh fuel from the research reactor to Kozloduy due to security reasons practically NPP Kozloduy remains the only significant (from safeguards point of view) nuclear site in Bulgaria. In 1972 a 'Nuclear Fuel' group was formed at the Physicists Department in NPP Kozloduy with responsibilities to carry out for safeguards records and reports, fresh and spent fuel transport and control. In 1990 this group was transferred to the Safety Section and since 1992 it exists as 'Control and Accounting for of the Nuclear Materials' - a section in the Safety Department. Currently the section serves all four facilities in NPP Kozloduy and has four people: section head, chief inspector and two inspectors. The main activities of the section include: a) Control of the nuclear fuel location as well as meeting the storage and transport conditions regulations; b) Control of the conditions for normal operations of the installed IAEA surveillance systems; c) Preparation of documents for licensing of fresh and spent nuclear fuel transport; d) Preparation of the official information on nuclear materials location and quantity; e) Preparation of accounting records and the reports for IAEA (ICR, PIL, MBR); f) Co-ordination of the IAEA safeguards inspection activities at NPP

  2. Reactor plant for Belene NPP completion

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Ryzhov, S. B.; Ermakov, D. N.; Repin, A. I.

    2004-01-01

    Construction of 'Belene' NPP was started at the end of 80-ties using project U-87 with V-320 reactor plant, general designer of this plant is OKB 'Gidropress'. At the beginning of 90-ties, on completing the considerable number of deliveries and performance of civil engineering work at the site the NPP construction was suspended. Nowadays, considering the state of affairs at the site and the work performed by Bulgarian Party on preservation of the equipment delivered, the most perspective is supposed to be implementation of the following versions in completing 'Belene' NPP: for completion of Unit 1 - reactor plant VVER-1000 on the basis of V-320 reactor with the maximum use of the delivered equipment (V-320M) having the extended service life and safety improvement; for Unit 2 - advanced reactor plant VVER-1000. For the upgraded reactor plant V-230M the basic solutions and characteristics are presented, as well as the calculated justification of strength and safety analyses, design of the reactor core and fuel cycle, instrumentation and control systems, application of the 'leak-before break' in the project and implementation of safety measures. For the modernised reactor plant V-392M the main characteristics and basic changes are presented, concerning reactor pressure vessel, steam generator, reactor coolant pump set. Design of NPP with the modernized reactor plant V-320M meets the up-to-date requirements and can be licensed for completion and operation. In the design of NPP with the advanced reactor plant the basic solutions and the equipment are used that are similar to those used in standard reactor plant V-320 and new one with VVER-1000 under construction and completion in Russia, and abroad. Compliance of reactor design with the up-to-date international requirements, considering the extended service life of the main equipment, shows its rather high potential for implementation during completion of 'Belene' NPP

  3. Overview of Cernavoda NPP training programme

    International Nuclear Information System (INIS)

    Valache, Cornelia

    2005-01-01

    Full text: The paper presents a general assessment of Cernavoda NPP personnel training programme, highlighting the role of training in human performance improvement. Cernavoda NPP Personnel Training and Authorization Department (PTAD) is the one responsible for the training of Unit 1 and 2 staff. PTAD is structured in such a manner as to support and respond to all Cernavoda NPP training, qualification and authorization requirements. The training of personnel is continuously adapted based on IAEA Guides and INPO/WANO recommendations, to keep with world standards, based on the internal and external reviews. At Cernavoda NPP the Training Concept and the Training Programmes are based on SAT - Systematic Approach to Training. The Training Concept is established on a set of training documents (RD's, SI's, IDP's), which address all the SAT phases: Analysis, Design, Development, Implementation and Evaluation. The Training Programmes are structured on the initial and continuing personnel training. Their content and goals respond to the training specific needs for each plant major job family. In order to successfully support NPP training programmes, CNPP training center has upgraded classrooms with new presentation facilities and there are plans to expand the space of the building, to develop additional operator and maintenance skills facilities. There are also a number of general areas for improvement of PTAD activity, such as operators and maintenance training, initial and continuing training of both Unit 1 and 2 staff. By responding in a timely and completely manner to all plant training requirements PTAD will help in rising human performance of Cernavoda NPP personnel, supporting the safe, efficient and cost effective production of power. (author)

  4. Perspectives of NPP personnel training in the future?

    International Nuclear Information System (INIS)

    Khess, R.E.

    1996-01-01

    Problems of the NPP personnel education are discussed during radical reorganization of the state regulation of energy production and distribution when nuclear energetics will appears in the market conditions. Effective methods of the NPP personnel training developed by the corporation GPU Nuclear are given. Potentials of the application of simulators and computerized programs for increasing the NPP personnel education efficiency are considered

  5. Organization of monitoring of agricultural products in NPP region

    International Nuclear Information System (INIS)

    Panteleev, L.I.; Spirin, E.V.; Sanzharova, N.I.

    1990-01-01

    Problem of organizing chemical and radiation monitoring of agricultural products in NPP region is considered. Attention is paid to monitoring during NPP siting and designing, to monitoring of radioactive contamination of agricultural products under normal NPP operation, emergency situations and decommissioning

  6. Calculation of isotope burn-up and change in efficiency of absorbing elements of WWER-1000 control and protection system during burn-up

    International Nuclear Information System (INIS)

    Timofeeva, O.A.; Kurakin, K.U.

    2006-01-01

    The report deals with fast and thermal neutron flows distribution in structural elements of WWER-1000 fuel assembly and absorbing rods, determination of absorbing isotope burn-up and worth variation in WWER reactor control and protection system rods. Simulation of absorber rod burn-up is provided using code package SAPPHIRE 9 5 end RC W WER allowing detailed description of the core segment spatial model. Maximum burn-up of absorbing rods and respective worth variation of control and protection system rods is determined on the basis of a number of calculations considering known characteristics of fuel cycles (Authors)

  7. Current status of Chernobyl NPP decommissioning

    International Nuclear Information System (INIS)

    2009-01-01

    Strategy of Chernobyl NPP decommissioning with the decommissioning license 2002-2064 is presented. The main activities at the stage of ChNPP units shutdown (2002 - 2012) are: units maintenance in safe state; decommissioning infrastructure construction; unloading of SNF – main activity determining the stage duration; systems and elements final shutdown; decommissioning life-support systems reconstruction; Comprehensive engineering and radiation survey (CERS); dismantling of the reactor facilities external equipment; removal of RAW from units; decommissioning documentation development. The decommissioning activities main results are presented

  8. Radioactive waste problems in the Kozloduy NPP

    International Nuclear Information System (INIS)

    Videnov, N.; Stanchev, V.

    1995-01-01

    An average volume of 1400 m 3 a year of solid radioactive waste (RAW) is generated in the Kozloduy NPP. The adopted waste processing sequence is collection, sorting and compaction with a 1000 tons force providing decrease in volume by factor of 15. A temporary storage facility at the Kozloduy NPP is licensed by ISUAE and CPPUAE. The treatment of liquid wastes is performed by Westinghouse formula and a technology using an automated solidification system. Contaminated oils are burned using an oil incinerator. A special 2-year programme for RAW management is being developed

  9. NPP Krsko natural circulation performance evaluation

    International Nuclear Information System (INIS)

    Segon, Velimir; Bajs, Tomislav; Frogheri, Monica

    1999-01-01

    The present document deals with an evaluation of the natural circulation performance of the Krsko nuclear power plant. Two calculation have been performed using the NPP Krsko nodalization (both similar to the LOBI A2-77 natural circulation experiment) - the first with the present steam generators at NPP Krsko (Westinghouse, 18% plugged), the second with the future steam generators (Siemens, 0% plugged). The results were evaluated using the natural circulation flow map derived in /1/, and were compared to evaluate the influence of the new steam generators on the natural circulation performance. (author)

  10. Knowledge management during decommissioning of Chornobyl NPP

    International Nuclear Information System (INIS)

    Gora, A.D.; Kuchinskij, V.K.; Stel'makh, D.A.; Tsivun, T.A.

    2010-01-01

    The article deals with issues on knowledge management during decommissioning by the example of the Chornobyl NPP. This includes how the duration of decommissioning stage, change in organization goal and final state of the site influence on human resources and knowledge management system. The main attention is focused on human assets and intellectual strength of Chornobyl NPP. Mathematical dependencies are proposed to substantiate numerical values. An analysis is given for the current situation, and forecast estimates for values dynamics is performed. The conclusion gives solutions on providing experienced staff in the future.

  11. Radioactive waste problems in the Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Videnov, N; Stanchev, V [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    An average volume of 1400 m{sup 3} a year of solid radioactive waste (RAW) is generated in the Kozloduy NPP. The adopted waste processing sequence is collection, sorting and compaction with a 1000 tons force providing decrease in volume by factor of 15. A temporary storage facility at the Kozloduy NPP is licensed by ISUAE and CPPUAE. The treatment of liquid wastes is performed by Westinghouse formula and a technology using an automated solidification system. Contaminated oils are burned using an oil incinerator. A special 2-year programme for RAW management is being developed.

  12. Medical consequences of NPP and TPP operation

    International Nuclear Information System (INIS)

    Bliznakov, V.

    1996-01-01

    Results from a comparative analysis of health conditions of the staff in the Kozloduy NPP and the Maritsa Iztok TPP are reported. It is found that the general disease incidence with temporary incapacity for work of Kozloduy workers is lower than those data for the workers at thermal power stations. The incidence of some social diseases like neoplasms, TBC, hypertension, ischemia etc. is also lower for the staff of NPP. No cases of radiation injuries have been registered for a period of 21 years

  13. Safety upgrading program in NPP Mochovce

    International Nuclear Information System (INIS)

    Baumeister, P.

    1999-01-01

    EMO interest is to operate only nuclear power plants with high standards of nuclear safety. This aim EMO declare on preparation completion and commissioning of Mochovce Nuclear Power Plant. Wide co-operation of our company with International Atomic Energy Agency and west European Inst.ions and companies has been started with aim to fulfil the nuclear safety requirements for Mochovce NPP. Set of 87 safety measures was implemented at Mochovce Unit 1 and is under construction at Unit 2. Mochovce NPP approach to safety upgrading implementation is showed on chosen measures. This presentation is focused on the issues category III.(author)

  14. Lifetime evaluation of Bohunice NPP components

    International Nuclear Information System (INIS)

    Kupca, L.

    2001-01-01

    The paper discuss some aspects of the main primary components lifetime evaluation program in Bohunice NPP which is performed by Nuclear Power Plant Research Institute (NPPRI) Trnava in cooperation with Bohunice and other organizations involved. Facts presented here are based on the NPPRI research report which is regularly issued after each reactor fuel campaign under conditions of project resulted from the contract between NPPRI and Bohunice NPP. For the calculations, there has been used some computer codes adapted (or made) by NPPRI and the results are just the conclusive and very brief, presented here in Tables (Figures). (authors)

  15. NPP Krsko small break LOCA analysis

    International Nuclear Information System (INIS)

    Mavko, B.; Petelin, S.; Peterlin, G.

    1987-01-01

    Parametric analysis of small break loss of coolant accident for the Krsko NPP was calculated by using RELAP5/MOD1 computer code. The model that was used in our calculations has been improved over several years and was previously tested in simulation (s) of start-up tests and known NPP Krsko transients. In our calculations we modelled automatic actions initiated by control, safety and protection systems. We also modelled the required operator actions as specified in emergency operating instructions. In small-break LOCA calculations, we varied break sizes in the cold leg. The influence of steam generator tube plugging on small break LOCA accidents was also analysed. (author)

  16. Implementation of burnup in FERM nodal computer code

    International Nuclear Information System (INIS)

    Yoriyaz, H.; Nakata, H.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  18. Development and verification of Monte Carlo burnup calculation system

    International Nuclear Information System (INIS)

    Ando, Yoshihira; Yoshioka, Kenichi; Mitsuhashi, Ishi; Sakurada, Koichi; Sakurai, Shungo

    2003-01-01

    Monte Carlo burnup calculation code system has been developed to evaluate accurate various quantities required in the backend field. From the Actinide Research in a Nuclear Element (ARIANE) program, by using, the measured nuclide compositions of fuel rods in the fuel assemblies irradiated in the commercial Netherlands BWR, the analyses have been performed for the code system verification. The code system developed in this paper has been verified through analysis for MOX and UO2 fuel rods. This system enables to reduce large margin assumed in the present criticality analysis for LWR spent fuels. (J.P.N.)

  19. Burn-up TRIGA Mark II benchmark experiment

    International Nuclear Information System (INIS)

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

    1998-01-01

    Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

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

    International Nuclear Information System (INIS)

    1998-04-01

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

  1. Preparation of computer codes for analyzing sensitivity coefficients of burnup characteristics (2) (Contract research, translated document)

    International Nuclear Information System (INIS)

    Hanaki, Hiroshi; Sanda, Toshio; Ohashi, Masahisa

    2008-10-01

    To develop nuclear design of LMFBR cores, they are important subjects of research and development to improve the accuracy in nuclear design of large LMFBR cores and to design highly efficient core more rationally. The adjusted nuclear cross-sections library has been made by being reflected the result of critical experiment of the JUPITER, etc. effectively as much as possible. And the distinct improvement of the accuracy in nuclear design of large LMFBR cores has been achieved. In the design of large LMFBR cores, however, it is important to accurately estimate not only nuclear characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. Therefore, it is thought to improve the prediction accuracy for burnup characteristics using many burnup data of 'Joyo' effectively. It is thought the best way to adjust cross sections using sensitivity coefficients of burnup characteristics to utilize burnup data of 'Joyo'. It is able to know the accuracy quantitatively for burnup characteristics of large LMFBR by analyzing the sensitivity coefficients. Therefore in this work computer codes for analyzing sensitivity coefficients of burnup characteristics had been prepared since 1992. In 1992 cross-section adjustment was done by using the data of 'Joyo' and the effect was studied. In this year the adequacy of the codes was studied with a view of applying of design of large LMFBR cores. The results are as follows: (1) The computer codes which could analyze sensitivity coefficients of burnup characteristics taking into consideration plural cycles and refueling were prepared, therefore it came of be able to adjust cross sections using burnup data and to estimate the accuracy for design of large LMFBR cores. The characteristics are not only burnup reactivity loss, breeding ratio but also number density, criticality, reactivity worth, reaction rate ratio, and reaction rate

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-01

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

  3. External flood probabilistic safety analysis of a coastal NPP

    International Nuclear Information System (INIS)

    Pisharady, Ajai S.; Chakraborty, M.K.; Acharya, Sourav; Roshan, A.D.; Bishnoi, L.R.

    2015-01-01

    External events pose a definitive challenge to safety of NPP, solely due to their ability to induce common cause failures. Flooding incidents at Le Blayais NPP, France, Fort Calhoun NPP, USA and Fukushima Daiichi have pointed to the importance of external flooding as an important contributor to NPP risk. A methodology developed for external flood PSA of a coastal NPP vulnerable to flooding due to tsunami, cyclonic storm and intense local precipitation is presented in this paper. Different tasks for EFPSA has been identified along with general approach for completing each task

  4. Principles of tariff determination for NPP electric power generation

    International Nuclear Information System (INIS)

    Ratnikov, B.E.; Gitel'man, L.D.; Artemov, Yu.N.; Fiantsev, V.S.

    1988-01-01

    Foundations of price-setting and order of accounting arrangement for NPP electric power are considered. NPP tariffs are established proceeding from standard costs of power generation. The standards are differentiated as to NPP groups, depending on technical, regional and natural geographic factors, taking into account the facility type, unit capacity and the number of similar NPP units. The conclusion is made that under conditions of NPP economic independence expansion and creation of prerequisites for going over to self-financing principles and also due to the qualitatively new stage of nuclear power generation development the level of efficiency, forseen by the tariffs, should be increased

  5. Computer programs for TRIGA calibration, burnup evaluation, and bookkeeping

    International Nuclear Information System (INIS)

    Nelson, George W.

    1978-01-01

    Several computer programs have been developed at the University of Arizona to assist the direction and operation of the TRIGA Reactor Laboratory. The programs fall into the following three categories: 1. Programs for calculation of burnup of each fuel element in the reactor core, for maintaining an inventory of fuel element location and fissile content at any time, and for evaluation of the reactivity effects of burnup or proposed fuel element rearrangement in the core. 2. Programs for evaluation, function fitting, and tabulation of control rod measurements. 3. Bookkeeping programs to summarize and tabulate reactor runs and irradiations according to time, energy release, purpose, responsible party, etc. These summarized data are reported in an annual operating report for the facility. The use of these programs has saved innumerable hours of repetitious work, assuring more accurate, objective results, and requiring a minimum of effort to repeat calculations when input data are modified. The programs are written in FORTRAN-IV, and have been used on a CDC-6400 computer. (author)

  6. chemical determination of burnup ratio in nuclear fuels

    International Nuclear Information System (INIS)

    Guereli, L.

    1997-01-01

    Measurements of the extent of fission are important to determine the irradiation performance of a nuclear fuel. The energy released per unit mass of uranium (burnup) can be determined from measurement of the percent of heavy atoms that have fissioned during irradiation.The preferred method for this determination is choosing a suitable fission monitor (usually ''1''4''8Nd) and its determination after separation from the fuel matrix. In thermal reactor fuels where the only heavy element in the starting material is uranium, uranium depletion can be used for burnup determination. ''2''3''5U depletion method requires measurement of uranium isotopic ratios of both irradiated and unirradiated fuel. Isotopic ratios can be determined by thermal ionization mass spectrometer following separation of uranium from the fuel matrix. Separation procedures include solvent extraction, ion exchange and anion exchange chromatography. Another fission monitor used is ''1''3''9La determination by HPLC. Because La is monoisotopic (''1''3''9La) in the fuel, it can be determined by chemical analysis techniques

  7. Higher order methods for burnup calculations with Bateman solutions

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  8. Akkuyu NPP – the first Turkish NPP. The new history of the project

    International Nuclear Information System (INIS)

    Tzocheva, V.

    2012-01-01

    An overview is given to the Turkish energy sector and nuclear power plans. The project for the construction of the first NPP in Turkey is presented. The general parameters of the Project are: CAPEX: $ 20 bln; Project design: NPP-2006; (VVER- 1200); Number of units: 4; Total capacity: 4 800 MW; Construction period: 2014 – 2023; PPA period; 15 years, fixed price terms. An account of the activities during 2011, the Worley Parsons participation are presented and a tentative project schedule is given

  9. Techniques and results of examination of fission product release from VVER fuel rods with artificial defects and a burnup of ∼60 MWd/kgU at the MIR loop facility

    International Nuclear Information System (INIS)

    Burukin, A.; Goryachev, A.; Ilyenko, S.; Izhutov, A.; Konyashov, V.; Shishin, V.; Shulimov, V.; Luzanova, L.; Miglo, V.

    2009-01-01

    Complex of equipment and several techniques for examination of radioactive fission product release from defective fuel rods were developed, prepared and tested at the PV-1 loop facility of the MIR reactor. During the first test, which was conducted at the PV-1 loop facility and aimed at testing of developed equipment and techniques, measurement of radioactive fission product release from an experimental re-fabricated fuel rod with a burnup of ∼60 MWd/kgU and an artificial defect was performed under design-basis steady-state operating conditions of the VVER-1000 reactor. PIE of all main parameters of the experimental defective fuel rod did not reveal any state peculiarities which could be caused by the artificial defect, i.e. fuel and cladding characteristics in the defect area did not differ from the initial ones (before testing) as well as their characteristics in areas distant from the defect; they are typical for fuel rods with a similar irradiation history in the VVER NPP. The gap in the experimental fuel rod was bridged due to close contact between fuel and cladding at increased fuel burnup; it can appreciable reduce release of radioactive fission products into the PV-1 primary coolant. This suggestion and quantitative characteristics of effect of gap bridging in a high-burnup fuel rod on radioactive fission product release should be investigated during the next tests performed at the PV-1 loop facility. Values of radioactive fission product release measured during the first test at the PV-1 loop facility in the MIR reactor will be used for development of an empirical engineering model in order to take into account high burnup effects and their impact on fission product release from fuel and defective fuel rods

  10. Radioactive source management in Daya Bay NPP

    International Nuclear Information System (INIS)

    Mao Chun Yang

    2000-01-01

    'Small sources causes big accidents' had occurred worldwide many times. Radioactive source management in Nuclear Power Plant in very important for its safety record. This paper introduces the way and experience of radioactive source management in Daya Bay NPP from aspects of clarifying the responsibilities, centralizing the management of high radioactivity sources, work process management and experience feedback etc. (author)

  11. Safety upgrading of Bohunice V1 NPP

    International Nuclear Information System (INIS)

    2001-01-01

    This CD is multimedia presentation of programme safety upgrading of Bohunice V1 NPP. It consist of next chapters: (1) Introductory speeches; (2) Nuclear power plant WWER 440; (3) Safety improvement; (4) Bohunice Nuclear power plants subsidiary; (5) Siemens; (6) REKON; (7) VUJE Trnava, Inc. - Engineering, Design and Research Organisation; (8) Album

  12. NPP Mochovce - a project of extraordinay significance

    International Nuclear Information System (INIS)

    Chwolik, I.; Debru, M.

    2000-01-01

    In this paper and in this presentation the reactor safety upgrading of two blocks of the NPP V-1 Bohunice, some results of participation on safety upgrading by the German-French consortium EUCOM (Framatome and Siemens-KWU) are presented. (author)

  13. Jose Cabrera NPP severe accident management activities

    International Nuclear Information System (INIS)

    Blanco, J.; Almeida, P.; Saiz, J.; Sastre, J.L.; Delgado, R.

    1998-01-01

    To prepare a common acting plan with respect to Severe Accident Management, in 1994 was founded the severe accident management ''ad-hoc'' working group from the Spanish Westinghouse PWR Nuclear Power Plant Owners Group. In this group actively collaborated the Jose Cabrera NPP Training Centre and the Department of Nuclear Engineering of UNION FENOSA. From this moment, Jose Cabrera NPP began the planning of its specific Severe Accident Management Program, which main point are Severe Accident Management Guidelines (SAMG). To elaborate this guidelines, the Spanish translation of Westinghouse Owners Group (WOG) Severe Accident Management Guidelines were considered the reference documents. The implementation of this Guidelines to Jose Cabrera NPP started on January 1997. Once the specific guidelines have been implemented to the plant, training activities for the personnel involved in severe accident issues will be developed. To prepare the training exercises MAAP4 code will be used, and with this intention, a specific Jose Cabrera NPP MAAP-GRAAPH screen has been developed. Furthermore, a wide selection of MAAP input files for the simulation of different scenarios and accidental events is available. (Author)

  14. External hazards considered for Paks NPP

    International Nuclear Information System (INIS)

    Kiss, Tibor

    2000-01-01

    PAKS NPP was built according to Soviet construction standards which took into account meteorological aspects but no documents for other external hazards were available. Main activities concerning earthquakes cover reevaluation of the plant site, seismic safety technological concept, improving the seismic resistance, installation of seismic monitoring and protection system, and seismic PSA

  15. Maintenance training centre at NPP Paks, Hungary

    International Nuclear Information System (INIS)

    Babos, K.

    1996-01-01

    The lecture shows the feature of WWER-440/213 units maintenance, the existing maintenance training system, the necessity of the change in maintenance training system at NPP Paks. The author introduces the would-be maintenance training centre, the training facilities and the main tasks related to the maintenance training. (author)

  16. Safety culture development at Daya Bay NPP

    International Nuclear Information System (INIS)

    Zhang Shanming

    2001-01-01

    From view on Organization Behavior theory, the concept, development and affecting factors of safety culture are introduced. The focuses are on the establishment, development and management practice for safety culture at Daya Bay NPP. A strong safety culture, also demonstrated, has contributed greatly to improving performance at Daya Bay

  17. Operational experiences feedback in Bohunice NPP

    Energy Technology Data Exchange (ETDEWEB)

    Betak, A [NPP Bohunice (Slovakia)

    1997-10-01

    The presentation reviews the following issues: OEF team in Bohunice NPP - structure; training and qualification: ASSET seminars on Prevention of incidents - INES manual handling, NRA-NRC the training on event investigation methods, NU - the training on HPES; legislation - documentation prepared in the frame of QA programme; results of OEF team activities; ASSET mission Dukovany - Experiences; the perspective activities.

  18. Development of the NPP Kozloduy informational system

    Energy Technology Data Exchange (ETDEWEB)

    Katsarov, Y; Manchev, B [Risk Engineering Ltd., Sofia (Bulgaria)

    1996-12-31

    A unified computer system to be implemented in the Kozloduy NPP is described. It allows to generate all general indicators and documents for plant operation needed according to the national regulation and WANO Performance Indicators. The system combines multi-purpose data collection with probabilistic analysis, evaluation of the human factor and optimization of the regulated verification of the equipment. 3 refs.

  19. NPP radioactive waste processing and solidification

    International Nuclear Information System (INIS)

    Nikiforov, A.S.; Polyakov, A.S.; Zakharova, K.P.

    1983-01-01

    The problems of proce-sing NPP intermediate level- and low-level liquid radioactive wastes (LRW) are considered. Various methods are compared of LWR solidification on the base of bituminization, cement grouting and inclusion into synthetic resins. It is concluded that the considered methods ensure radioactive radionuclides effluents into open hydronetwork at the level below the sanitary, standards

  20. Biotic elements of NPP techno-ecosystem

    International Nuclear Information System (INIS)

    Protasov, A.A.; Silaeva, A.A.

    2013-01-01

    Specific features of biotic elements in the NPP techno-ecosystems were considered and compared with natural ecosystems. Relationships between biotic communities and environmental factors that are specific to the techno-ecosystems were discussed, and the problems of limitation of biological hindrances in operation of equipment, principles of hydrobiological and environmental monitoring were considered.

  1. Slovakia: Mochovce NPP. Project control. Annex 10

    International Nuclear Information System (INIS)

    Bodorik, K.

    1999-01-01

    This annex deals with project control. Mochovce NPP suffered considerable delay primarily due to lack of money. This situation was corrected and construction resumed in 1996. Throughout the 'dormant' period the plant received considerable support from the major contractors, who maintained skeleton staff at site. Significant safety and managerial improvements are being introduced and a strategic plan for the plant has been developed. (author)

  2. Development of NPP safety regulation in Russia

    International Nuclear Information System (INIS)

    Vishnevsky, Y.G.; Gutsalov, A.T.; Bukrinsky, A.M.; Gordon, B.G.

    1999-01-01

    The presentation describes the organisation scheme of Russian safety regulatory bodies, their tasks and responsibilities. Legislative and regulatory basis of NPP safety regulations rely on the federal laws: Law on the Use of Nuclear Energy and Law on Radiation Safety of the Population. Role of international cooperation and Improvement of regulatory activities in Russia are emphasised

  3. Determination of the burn-up of TRIGA fuel elements by calculation with new TRIGLAV program

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.

    1996-01-01

    The results of fuel element burn-up calculations with new TRIGLAV program are presented. TRIGLAV program uses two dimensional model. Results of calculation are compared to results calculated with program, which uses one dimensional model. The results of fuel element burn-up measurements with reactivity method are presented and compared with the calculated results. (author)

  4. Application of reactivity method to MTR fuel burn-up measurement

    International Nuclear Information System (INIS)

    Zuniga, A.; Ravnik, M.; Cuya, R.

    2001-01-01

    Fuel element burn-up has been measured for the first time by reactivity method in a MTR reactor. The measurement was performed in RP-10 reactor of Peruvian Institute for Nuclear Energy (IPEN) in Lima. It is a pool type 10MW material testing reactor using standard 20% enriched uranium plate type fuel elements. A fresh element and an element with well defined burn-up were selected as reference elements. Several elements in the core were selected for burn-up measurement. Each of them was replaced in its original position by both reference elements. Change in excess reactivity was measured using control rod calibration curve. The burn-up reactivity worth of fuel elements was plotted as a function of their calculated burnup. Corrected burn-up values of the measured fuel elements were calculated using the fitting function at experimental reactivity for all elements. Good agreement between measured and calculated burn-up values was observed indicating that the reactivity method can be successfully applied also to MTR fuel element burn-up determination.(author)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  6. 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)

  7. Determination of nuclear fuel burn-up using mass spectrometric techniques

    International Nuclear Information System (INIS)

    Saha, B.; Bagyalakshmi, R.; Periaswami, G.; Kavimandan, V.D.; Chitambar, S.A.; Jain, H.C.; Mathews, C.K.

    1977-01-01

    Determination of burn-up using a stable fission product monitor such as 148 Nd and heavy elements, determined by isotope dilution mass spectrometry gives the most accurate data. This report describes the work carried out to standardise the conditions for burn-up determination. Some typical results are given. (author)

  8. The application of burnup credit for spent fuel operations in the United Kingdom

    International Nuclear Information System (INIS)

    Bowden, R.

    1998-01-01

    This paper begins by outlining the structure of the nuclear industry in the United Kingdom. It then sets out the methodology of burnup credit, and provides a brief discussion of the validation and robustness of the calculational route. This leads to a description of both the current and intended applications of burnup credit in the United Kingdom. (author)

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

    Science.gov (United States)

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

    1992-06-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  11. Simulation of High Burnup Structure in UO2 Using Potts Model

    International Nuclear Information System (INIS)

    Oh, Jae Yong; Koo, Yang Hyun; Lee, Byung Ho

    2009-01-01

    The evolution of a high burnup structure (HBS) in a light water reactor (LWR) UO 2 fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the UO 2 matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels

  12. Sitting Safety Aspects of Second Romanian NPP

    International Nuclear Information System (INIS)

    Mauna, T.

    2010-01-01

    The first Romanian NPP CANDU 6 type reactor gone to erection in 1980 on Cernavoda site planned to have 5 units like the Wolsong applied design project for nuclear island. For the BOP parts the ASALDO-GE project was applied with the careful about the interface connection NSP requirements. The new NPP sitting studies began from 1982 in a serious manner as first part on Nuclear Power Plant Romanian Program adopted by political and governmental authorities at the time. For develop the all package of the studies in concordance with the first IAEA Safety Standards recommendations. Till the 1982 the first mission of design and research multi-branch of specialists team was to adapt the NPP Cernavoda project having a open water cooling circuit to the new parameters of close water cooling circuit. But the team was looking at the other type of NPP for sitting. Also in the same time was studied the possibility of NSP foundation on hard less or soft soil foundation strata in connection with safety aspects. The close circuit of cooling water means others parameters of systems and need very large cooling towers. Also must be reconsidering the safety systems design and performance as new solution. In the south of Transylvania historical region in Romania the Olt River run from west to east having medium multi annual flow around 70 m3/s. The Olt River has a chain of small hydropower in operation and other planned. From geological and geophysical points of view two main faults, along the Olt river valley, one of this having seismically small activities was detected. Site region geotechnical studies show small quantity underground natural gas, salt and peat. The initial nuclear program has imposed 4 NPP units site near Olt River. Taking into account the orogenesis, water cooling needs and other local feature can't be built more than two NPP units on a site. This paper tries to reconsider the old analysis from the last IAEA Safety Standards point of view taking into account the new

  13. Considerations related to Cernavoda NPP lifetime management

    International Nuclear Information System (INIS)

    Cojan, Mihail

    2007-01-01

    Cernavoda NPP, the first CANDU in Europe, is one of the original CANDU 6 plants and the first CANDU 6 producing over 706 MWe. While the first series of CANDU 6 plants (which entered service in the early 1980 s) have now reached the 2/3 of their 30 years design life, the Cernavoda NPP was put into service on the 2nd December 1996. After 10 years of operation the Plant Life Management (PLiM) Program for Cernavoda should be an increasingly important program to Utility ('CNE - Prod') in order to protect the investment and the continued success of plant operation. The goal of the paper is to present some considerations related to Cernavoda NPP lifetime management. The Plant Life Management Program, known as PLiM Program is concerned with the analysis of technical limits of the safe operation - from the point of view of nuclear safety - in NPP units, aiming at attaining the planned 30 years life duration and its extension to 40 or even 50 years of safe and economical operation. For the CANDU reactors the so-called PLiM and PLEX Programs are just applied. These are applied research programs that approach with priority the current practices for assessing the capability of safe operation within the limits of nuclear safety (fitness-for-service assessment). These programs also approach inspection, monitoring are prevention of degrading due to the ageing of critical systems, structures and components (CSSCs). As each nuclear plant is somewhat different in its components and systems, materials composition, procurement, construction, and operational history, directed research and development programs into materials behavior, monitoring techniques, and methods to mitigate ageing are required to support the lifetime management. Over the past 6 years, INR Pitesti (Institute for Nuclear Research - Romania) has been working on R and D Programs to support a comprehensive and integrated Cernavoda NPP Life Management Program (PLiM) that will see the Cernavoda NPP successfully and

  14. Safety Analysis Report for Ignalina NPP

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1997-01-01

    In December 1994 an agreement was signed between the European Bank for Reconstruction and Development and the Republic of Lithuania for the grant of 32.86 MECU for the safety Improvement at Ignalina NPP. One of the conditions for the provision of the grant, was a requirement for an in-depth analysis of the safety level at Ignalina NPP in the scope and according to the standards acceptable for a western nuclear power plant, and to publish a Safety Analysis Report (SAR). The report should investigate and analyze any factor that could limit a safe operation of the plant, and provide recommendations for actual safety improvements. According to the agreement, Lithuania had to finalize the SAR until 31 December, 1995. The bank has also organized and financed investigation of safety at Ignalina NPP and preparation of the SAR. EBRD made an agreement with Sweden's Vattenfall, which subcontracted well-known companies from Canada, USA, Germany, etc., and also the Russian Research and Development Institute of Power Engineering (NIKIET), reactor designer of Ignalina NPP. The SAR is a very comprehensive document and contains about 8000 pages of text, diagrams and tables. The main findings of the SAR are provided in the article. A large number of discrepancies with modern rules and western practices was detected, but they were not proved to be serious enough to require reactors shutdown. Based on the recommendations of the SAR Ignalina NPP has worked out Safety Improvement Program No. 2 (SIP-2), which is planned for three years and will cost 486 MLT. (author)

  15. Cernavoda NPP simulator - next generation

    International Nuclear Information System (INIS)

    Tatar, F.; Ionescu, T.; Dascalu, M.

    2003-01-01

    Demand for extending the amount of training and scope for Cernavoda Unit 1 as well as the new trend in the simulator owners world, led to a change in the Romanian philosophy of simulator specification. Up to now the training was conducted on a Full Scope simulator, a 1:1 replica of Cernavoda Unit 1 reference plant. The present task is to define the simulation facilities and structure capable to meet the requirements for training, qualification and licensing of personnel for both Cernavoda Unit 1 and Unit 2. Obviously, the Cernavoda Unit 2 belongs to the same technological family but has rather different control room layout. Since this target requires a new simulator the costs would be rather high in accordance to the degree of automation of Cernavoda NPP. Therefore, depending on training requirements and financing, the Cernavoda Unit 1 simulator modernization, which also provides an alternative to full scope control room simulator, may be a viable option. Therefore the solution that with discuss for Cernavoda training extension is the migration of Cernavoda Unit 1 simulator to state-of-the-art. Consequently, the Cernavoda Unit 1 simulator modernization task will be organized as project including the following major items: 1. Rehost existing U1 simulation software from VAX 4500 to: - Best commercial multi-processor server for simulation server (HP, O/S Linux); - Best commercial single processor PC for I/O communications (HP, O/S Linux); 2. Replace DCC with enhanced emulated version: Best commercial individual PC for DCC emulation (HP, O/S Windows); Support for actual keyboards; Replacement of RAMTEK System and CONRAC Monitors with X terminals or PC's; 3. Conversion of AutoCAD-based panel graphic pages to RAVE-based; 4. Install the required software tools for developing enhanced simulation modules; 5. Replace the simulation modules with advanced modules; 6. Replace the present Windows Instructor Facilities with ISIS; 7. Development of a selection of MCR-U1 virtual

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Jones, R.H.

    1989-01-01

    Sandia has examined the matter of burnup credit from the perspective of physics, logistics, risk, and economics. A limited survey of the nuclear industry has been conducted to get a feeling for the actual application of burnup credit. Based on this survey, it can be concluded that the suppliers of spent fuel storage and transport casks are in general agreement that burnup credit offers the potential for improvements in cask efficiency without increasing the risk of accidental criticality. The actual improvement is design-specific but limited applications have demonstrated that capacity increases in the neighborhood of 20 percent are not unrealistic. A number of these vendors acknowledge that burnup credit has not been reduced to practice in cask applications and suggest that operational considerations may be more important to regulatory acceptance than to the physics. Nevertheless, the importance of burnup credit to the nuclear industry as a cask design and analysis tool has been confirmed by this survey

  19. Regulatory status of burnup credit for storage and transport of spent fuel in Germany

    International Nuclear Information System (INIS)

    Neuber, J.C.; Schweer, H.H.; Johann, H.G.

    2001-01-01

    This paper describes the regulatory status of burnup credit applications to pond storage and dry-cask transport and storage of spent fuel in Germany. Burnup credit for wet storage of LWR fuel at nuclear power plants has to comply with the newly developed safety standard DIN 25471. This standard establishes the safety requirements for burnup credit criticality safety analysis of LWR fuel storage ponds and gives guidance on meeting these requirements. Licensing evaluations of dry transport systems are based on the application of the IAEA Safety Standards Series No.ST-1. However, because of the fact that burnup credit for dry-cask transport becomes more and more inevitable due to increasing initial enrichment of the fuel, and because of the increasing importance of dry-cask storage in Germany, the necessity of giving regulatory guidance on applying burnup credit to dry-cask transport and storage is seen. (author)

  20. Assessment of US NRC fuel rod behavior codes to extended burnup

    International Nuclear Information System (INIS)

    Laats, E.T.; Croucher, D.W.; Haggag, F.M.

    1982-01-01

    The purpose of this paper is to report the status of assessing the capabilities of the NRC fuel rod performance codes for calculating extended burnup rod behavior. As part of this effort, a large spectrum of fuel rod behavior phenomena was examined, and the phenomena deemed as being influential during extended burnup operation were identified. Then, the experiment data base addressing these identified phenomena was examined for availability and completeness at extended burnups. Calculational capabilities of the NRC's steady state FRAPCON-2 and transient FRAP-T6 fuel rod behavior codes were examined for each of the identified phenomenon. Parameters calculated by the codes were compared with the available data base, and judgments were made regarding model performance. Overall, the FRAPCON-2 code was found to be moderately well assessed to extended burnups, but the FRAP-T6 code cannot be adequately assessed until more transient high burnup data are available

  1. Cell verification of parallel burnup calculation program MCBMPI based on MPI

    International Nuclear Information System (INIS)

    Yang Wankui; Liu Yaoguang; Ma Jimin; Wang Guanbo; Yang Xin; She Ding

    2014-01-01

    The parallel burnup calculation program MCBMPI was developed. The program was modularized. The parallel MCNP5 program MCNP5MPI was employed as neutron transport calculation module. And a composite of three solution methods was used to solve burnup equation, i.e. matrix exponential technique, TTA analytical solution, and Gauss Seidel iteration. MPI parallel zone decomposition strategy was concluded in the program. The program system only consists of MCNP5MPI and burnup subroutine. The latter achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, and data exchanging with MCNP5MPI. Also, the program was verified with the pressurized water reactor (PWR) cell burnup benchmark. The results show that it,s capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)

  2. 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.

  3. Application of burnup credit in spent fuel management at Russian NPPs

    International Nuclear Information System (INIS)

    Koulikov, V.I.; Makarchuk, T.F.; Tikhonov, N.S.

    1998-01-01

    The article concerns implementation of burnup credit in spent fuel storage and transportation. Some of the problems with increased enrichment fuel can be resolved by use of modified transport methodology. Such as shipping in gas-filled casks only, reduced number of assemblies in casks, etc. However, the use of modified schemes of transportation results in essential financial losses. An actinide-only burnup credit is taken into account in most part of criticality calculations, and a parameter limiting loading of spent fuel in the cask or the repository is the avenge value of burnup on an assembly. The main method of burnup depth definition is its defect measurement. A short description of devices for measurement as well as some technical results of suing burnup credit approach in storage and transport are given. (author)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  5. Determination of burn-up of irradiated nuclear fuels using mass spectrometry

    International Nuclear Information System (INIS)

    Jagadish Kumar, S.; Telmore, V.M.; Shah, R.V.; Sasi Bhushan, K.; Paul, Sumana; Kumar, Pranaw; Rao, Radhika M.; Jaison, P.G.

    2017-01-01

    Burn-up defined as the atom percent fission, is a vital parameter used for assessing the performance of nuclear fuel during its irradiation in the reactor. Accurate data on the actinide isotopes are also essential for the reliable accountability of nuclear materials and for nuclear safeguards. Both destructive and non-destructive methods are employed in the post-irradiation analysis for the burn-up measurements. Though non-destructive methods are preferred from the point view of remote handling of irradiated fuels with high radioactivity, they do not provide the high accuracy as achieved by the chemical analysis methods. Thus destructive radiochemical and chemical analyses are still the established reference methods for accurate and reliable burn-up determination of irradiated nuclear fuels. In the destructive method, burn-up of irradiated nuclear fuel is determined by correlating the amount of a fission product formed during irradiation with that of heavy elements. Thus the destructive experimental determination of burn-up involves the dissolution of irradiated fuel samples followed by the separation and determination of heavy elements and fission product(s) to be used as burn-up monitor(s). Another approach for the experimental determination of burn-up is based on the changes in the abundances of the heavy element isotopes. A widely accepted method for burn-up determination is based on stable "1"4"8Nd and "1"3"9La as burn-up monitors. Several properties such as non-volatility, nearly same yields for thermal fissions of "2"3"5U and "2"3"9Pu etc justifies the selection of "1"4"8Nd as a burn-up monitor

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-06-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO 2 is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO 2 to higher oxides. The oxidation of UO 2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO 2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO 2 to UO 2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO 2.4 to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO 2 oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO 2 and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5)

  11. Revision of Krsko NPP Quality Assurance Plan

    International Nuclear Information System (INIS)

    Biscan, R.; Fifnja, I.; Kavsek, D.

    2012-01-01

    International standards from nuclear power plant operation area are being frequently upgraded and revised in accordance with the continuous improvement philosophy. This philosophy applies also to the area of Quality Assurance, which has also undergone significant improvement since the early 1950s. Besides just nuclear industry, there are also other international quality standards that are being continuously developed and revised, bringing needs for upgrades also in the nuclear application. Since the beginning of Krsko NPP construction, the overall Quality Assurance program and its applicable procedures were in place to assure that all planned and systematic actions necessary to provide adequate confidence that an item or service will satisfy given requirements to quality, are in place. The overall requirements for quality as one of the major objectives for Krsko NPP operation are also set forth in the Updated Safety Analyses Report, the document that serves as a base for operating license. During more than 30 years of Krsko NPP operation, the quality requirements and related documents were revised and upgraded in several attempts. The latest revision 6 of QD-1, Quality Assurance Plan was issued during the year 2011. The bases for the revision were: Changes of the Slovenian regulatory requirements (ZVISJV, JV5, JV9?), Changes of Krsko NPP licensing documents (USAR section 13?), SNSA inspection requirements, Changes of international standards (IAEA, ISO?), Conclusions of first PSR, Implementation of ISO standards in Krsko NPP (ISO14001, ISO17025), Changes of plant procedures, etc. One of the most obvious changes was the enlargement of the QA Plan scope to cover interdisciplinary areas defined in the plant management program MD-1, such as Safety culture, Self-assessment, Human performance, Industrial Safety etc. The attachment of the QA Plan defining relationships between certain standards was also updated to provide matrix for better correlation of requirements of

  12. 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.

  13. 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).

  14. Visualization of fuel rod burnup analysis by Scilab

    International Nuclear Information System (INIS)

    Tsai, Chiung-Wen

    2013-01-01

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

  15. Visualization of fuel rod burnup analysis by Scilab

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-15

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

  16. User's manual for the reactor burnup system, REBUS

    International Nuclear Information System (INIS)

    Olson, A.P.; Regis, J.P.; Meneley, D.A.; Hoover, L.J.

    1972-01-01

    A user's manual for the REBUS System (REactor BUrnup System) is presented. Its primary purpose is to provide sufficient information about the REBUS capability to the user to ensure its efficient utilization. The current REBUS System either solves for the infinite time (equilibrium) operating conditions of a recycle system under fixed conditions, or solves for operating conditions during a single time step (non-equilibrium). The capability of studying various in-reactor fuel management and ex-reactor fuel management schemes has been included. REBUS has been operated with one- and two-dimensional diffusion theory neutronics solutions up to the present time. The model was specifically designed for extension to other neutronics models such as three-dimensional diffusion or transport theory and direct or synthesis solutions

  17. Manufacturing Data Uncertainties Propagation Method in Burn-Up Problems

    Directory of Open Access Journals (Sweden)

    Thomas Frosio

    2017-01-01

    Full Text Available A nuclear data-based uncertainty propagation methodology is extended to enable propagation of manufacturing/technological data (TD uncertainties in a burn-up calculation problem, taking into account correlation terms between Boltzmann and Bateman terms. The methodology is applied to reactivity and power distributions in a Material Testing Reactor benchmark. Due to the inherent statistical behavior of manufacturing tolerances, Monte Carlo sampling method is used for determining output perturbations on integral quantities. A global sensitivity analysis (GSA is performed for each manufacturing parameter and allows identifying and ranking the influential parameters whose tolerances need to be better controlled. We show that the overall impact of some TD uncertainties, such as uranium enrichment, or fuel plate thickness, on the reactivity is negligible because the different core areas induce compensating effects on the global quantity. However, local quantities, such as power distributions, are strongly impacted by TD uncertainty propagations. For isotopic concentrations, no clear trends appear on the results.

  18. Validating analysis methodologies used in burnup credit criticality calculations

    International Nuclear Information System (INIS)

    Brady, M.C.; Napolitano, D.G.

    1992-01-01

    The concept of allowing reactivity credit for the depleted (or burned) state of pressurized water reactor fuel in the licensing of spent fuel facilities introduces a new challenge to members of the nuclear criticality community. The primary difference in this analysis approach is the technical ability to calculate spent fuel compositions (or inventories) and to predict their effect on the system multiplication factor. Isotopic prediction codes are used routinely for in-core physics calculations and the prediction of radiation source terms for both thermal and shielding analyses, but represent an innovation for criticality specialists. This paper discusses two methodologies currently being developed to specifically evaluate isotopic composition and reactivity for the burnup credit concept. A comprehensive approach to benchmarking and validating the methods is also presented. This approach involves the analysis of commercial reactor critical data, fuel storage critical experiments, chemical assay isotopic data, and numerical benchmark calculations

  19. Oxide thickness measurement for monitoring fuel performance at high burnup

    International Nuclear Information System (INIS)

    Jaeger, M.A.; Van Swam, L.F.P.; Brueck-Neufeld, K.

    1991-01-01

    For on-site monitoring of the fuel performance at high burnup, Advanced Nuclear Fuels uses the linear scan eddy current method to determine the oxide thickness of irradiated Zircaloy fuel cans. Direct digital data acquisition methods are employed to collect the data on magnetic storage media. This field-proven methodology allows oxide thickness measurements and rapid interpretation of the data during the reactor outages and makes it possible to immediately reinsert the assemblies for the next operating cycle. The accuracy of the poolside measurements and data acquisition/interpretation techniques have been verified through hot cell metallographic measurements of rods previously measured in the fuel pool. The accumulated data provide a valuable database against which oxide growth models have been benchmarked and allow for effective monitoring of fuel performance. (orig.) [de

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

    International Nuclear Information System (INIS)

    Schutte, C.K.

    1991-05-01

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

  1. Fuel performance at high burnup for water reactors

    International Nuclear Information System (INIS)

    1991-02-01

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

  2. Approach to lithium burn-up effect in lithium ceramics

    International Nuclear Information System (INIS)

    Rasneur, B.

    1994-01-01

    The lithium burn-up in Li 2 ZrO 3 is simulated by removing lithium under Li 2 O form and trapping it in high specific surface area powder while heating during 15 days or 1 month at moderate temperature so that lithium mobility be large enough without causing any sintering neither of the specimens nor of the powder. In a first treatment at 775 deg C during 1 month. 30% of the lithium content could be removed inducing a lithium concentration gradient in the specimen and the formation of a lithium-free monoclinic ZrO 2 skin. Improvements led to similar results at 650 deg C and 600 deg C, the latter temperatures are closer to the operating temperature of the ceramic breeder blanket of a fusion reactor. (author) 4 refs.; 4 figs.; 1 tab

  3. Time resolved measurements of triton burnup in JET plasmas

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Dieudonne, C.

    2013-01-01

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

  5. Restructuring of burnup sensitivity analysis code system by using an object-oriented design approach

    International Nuclear Information System (INIS)

    Kenji, Yokoyama; Makoto, Ishikawa; Masahiro, Tatsumi; Hideaki, Hyoudou

    2005-01-01

    A new burnup sensitivity analysis code system was developed with help from the object-oriented technique and written in Python language. It was confirmed that they are powerful to support complex numerical calculation procedure such as reactor burnup sensitivity analysis. The new burnup sensitivity analysis code system PSAGEP was restructured from a complicated old code system and reborn as a user-friendly code system which can calculate the sensitivity coefficients of the nuclear characteristics considering multicycle burnup effect based on the generalized perturbation theory (GPT). A new encapsulation framework for conventional codes written in Fortran was developed. This framework supported to restructure the software architecture of the old code system by hiding implementation details and allowed users of the new code system to easily calculate the burnup sensitivity coefficients. The framework can be applied to the other development projects since it is carefully designed to be independent from PSAGEP. Numerical results of the burnup sensitivity coefficient of a typical fast breeder reactor were given with components based on GPT and the multicycle burnup effects on the sensitivity coefficient were discussed. (authors)

  6. Full Core Burn-up Calculation at JRR-3 with MVP-BURN

    International Nuclear Information System (INIS)

    Komeda, Masao; Yamamoto, Kazuyoshi; Kusunoki, Tsuyoshi

    2008-01-01

    Research reactors use a burnable poison to suppress an excess reactivity in the beginning of reactor lifetime. The JRR-3 (Japan Research Reactor No.3) has used cadmium wires of radius 0.02 cm as a burnable poison. This report describes burn-up calculations of plate fuel models and full core models with MVP-BURN, which is a burn-up calculation code using Monte Carlo method and has been developed in JAEA (Japan Atomic Energy Agency). As the results of calculations of plate models, between a model composed of one burn-up region along the radius direction and a model composed of a few burn-up regions along the radius direction, the effective absorption cross section of 113 Cd has had different tendency on reaching approximate 40. day (10000 MWd/t). And as results of calculations of full core model, it has been indicated that k eff is almost same till approximate 80. day (22000 MWd/t) between a model composed of one burn-up region along the vertical direction and a model composed of a few burn-up regions along the vertical direction. However difference of 113 Cd burn-up becomes pronounced and each k eff makes a difference after 80. day. (authors)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  8. Analyzing the BWR rod drop accident in high-burnup cores

    International Nuclear Information System (INIS)

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

    1995-01-01

    This study was undertaken for the US Nuclear Regulatory Commission to determine the fuel enthalpy during a rod drop accident (RDA) for cores with high burnup fuel. The calculations were done with the RAMONA-4B code which models the core with 3-dimensional neutron kinetics and multiple parallel coolant channels. The calculations were done with a model for a BWR/4 with fuel bundles having burnups up to 30 GWd/t and also with a model with bundle burnups to 60 GWd/t. This paper also discusses potential sources of uncertainty in calculations with high burnup fuel. One source is the ''rim'' effect which is the extra large peaking of the power distribution at the surface of the pellet. This increases the uncertainty in reactor physics and heat conduction models that assume that the energy deposition has a less peaked spatial distribution. Two other sources of uncertainty are the result of the delayed neutron fraction decreasing with burnup and the positive moderator temperature feedback increasing with burnup. Since these effects tend to increase the severity of the event, an RDA calculation for high burnup fuel will underpredict the fuel enthalpy if the effects are not properly taken into account. Other sources of uncertainty that are important come from the initial conditions chosen for the RDA. This includes the initial control rod pattern as well as the initial thermal-hydraulic conditions

  9. Determination of enrichment of recycle uranium fuels for different burnup values

    International Nuclear Information System (INIS)

    Zabunoglu, Okan H.

    2008-01-01

    Uranium (U) recovered from spent LWR fuels by reprocessing, which contains small amounts of U-236, is to be enriched before being re-irradiated as the recycle U. During the enrichment of recovered U in U-235, the mass fraction of U-236 also increases. Since the existence of U-236 in the recycle U has a negative effect on neutron economy, a greater enrichment of U-235 in the recycle U is required for reaching the same burnup as can be reached by the fresh U fuel. Two burnup values play the most important role in determining the enrichment of recycle U: (1) discharge burnup of spent fuel from which the recycle U is obtained and (2) desired discharge burnup of the recycle U fuel. A step-by-step procedure for calculating the enrichment of the recycle U as a function of these two burnup values is introduced. The computer codes MONTEBURNS and ORIGEN-S are made use of and a three-component (U-235, U-236, U-238) enrichment scheme is applied for calculating the amount of U-236 in producing the recycle U from the recovered U. As was aimed, the resulting expression is simple enough for quick/hand calculations of the enrichment of the recycle U for any given discharge burnup of spent fuel and for any desired discharge burnup of the recycle U fuel, most accurately within the range of 33,000-50,000 MWd/tonU

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

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2001-01-01

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

  11. Modernization of the oldest Swedish NPP

    International Nuclear Information System (INIS)

    Hagberth, Ronald

    1998-01-01

    OKG operates three BWR units of ABB design: Oskarshamn 1 with a net capacity of 440' MW, Oskarshamn 2 of 600 MW and Oskarshamn 3 of 1160 MW. Oskarshamn 1 NPP was commissioned in 1972 as the first commercial nuclear unit in Sweden. After more than twenty years of successful operation, the unit is now also the first reactor in Sweden to undergo a large safety modernization program. In the year 2000 the Oskarshamn 1 NPP will be modernized to a high level of safety standard and ready for operation for another period of at least 20 years. Experience gained can be used when modernizing other NPPs. The investment program for life extension is reasonable and shows that NPPs can be operated with an expected life span of more than 40 years at an ever-increasing safety level and still be very competitive in a deregulated market. (author)

  12. Intelligent system for accident identification in NPP

    International Nuclear Information System (INIS)

    Hernandez, J.L.

    1998-01-01

    Accidental situations in NPP are great concern for operators, the facility, regulatory bodies and the environmental. This work proposes a design of intelligent system aimed to assist the operator in the process of decision making initiator events with higher relative contribution to the reactor core damage occur. The intelligent System uses the results of the pre-operational Probabilistic safety Assessment and the Thermal hydraulic Safety Analysis of the NPP Juragua as source for building its knowledge base. The nucleus of the system is presented as a design of an intelligent hybrid from the combination of the artificial intelligence techniques fuzzy logic and artificial neural networks. The system works with variables from the process of the first circuit, second circuit and the containment and it is presented as a model for the integration of safety analyses in the process of decision making by the operator when tackling with accidental situations

  13. Quality of Industry Support to NPP Krsko

    International Nuclear Information System (INIS)

    Nemcic, K.

    2008-01-01

    NPP Krsko developed program for Supplier evaluation and performance. During the regular control of suppliers and evaluation of industry support to NPP Krsko quality problems were reported. Different quality systems were evaluated and different suppliers as: design organizations, equipment manufacturers, material vendors were audited or surveillance was performed. This paper discuss and report various cases where quality issues were problems based on audit results and present actions and efforts undertaken by the NE Krsko Quality Assurance Department to improve performance of the contractors, vendors, suppliers. New and different quality standards as approach in numerous articles are described as improvement or quality changes but also 'different opinion exist'. This paper also presents the author view and approach how to solve the possible future problems with different quality systems and organisations used by industry who support daily operation of NE Krsko and give recommendations for future nuclear projects.(author)

  14. SAT for NPP personnel training in USA

    International Nuclear Information System (INIS)

    Spinney, R.

    1995-01-01

    This discussion addressed the experience with the application of SAT at USA NPPs. In particular, the transition of NPP training processes, staff composition, and reporting structure from the TMI accident to present. As well, oversight and guidance activities of the INPO and more intensive inspection by the NRC began during this period. The average NPP training staff grew to 30-40 per unit, along with a change in reporting line from plant to corporate management. With the reduction of resources occurring in the late 1980s, overall training staff size decreased, the composition changed, and reporting line reverted to plant management. The overall lessons-learned for application of the SAT consisted of the need for simplification, management involvement, and exploitation of the technology

  15. LTO License Application Project NPP Borssele

    International Nuclear Information System (INIS)

    Jong, A.E. de; Blom, F.J.; Leilich, J.

    2012-01-01

    Borssele NPP plans to extend its operating life with 20 years until 2034. Borssele has started the project LTO 'bewijsvoering' (LTO 'Justification') in order to meet the requirements of the Dutch regulator. The outline of the project is based on IAEA safety guide 57 'Safe Long Term Operation of Nuclear Power Plants'. This paper describes the contents and coherence of the different parts in the project and how these respond to the IAEA guidelines on LTO. The goal of the project LTO 'bewijsvoering' is to ensure that safety and safety relevant systems, structures and components continue to perform their intended functions during long term operation. The outcome of the project LTO 'bewijsvoering' will be used for a license change application and this will be submitted to the Dutch regulator KFD for approval of prolonged operation of Borssele NPP after 2013. (author)

  16. NPP Temelin. Status of safety improvements

    International Nuclear Information System (INIS)

    1999-01-01

    The WWER-1000 Temelin NPP under construction has been subjected as other NPPs of the same type to numerous project reviews resulting in quite a number of recommendations for design changes. Results of the IAEA mission to review the resolution of WWER-1000 safety issues at Temelin NPP are cited in this paper. The main conclusions emphasize that a combination of eastern and western technology and practices led to safety improvements in comparison with the international practices. Plant managers are clearly committed to implementation of operational programs which are consistent with effective western operational safety practices. Considerable effort remains to bring planned programs to successful implementation, in particular in meeting the need to foster strong safety culture among all personnel

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  18. Organizational aspects of NPP operator training

    International Nuclear Information System (INIS)

    Vel'chinskij, V.I.

    1992-01-01

    The main points of the document regulating the selection, prepation, permission for work and in-service control of NPP personnel developed on the basis of the IAEA requirements are considered. The specialists engaged for work are subjected to qualification, medical, professional, psychological and psychophysiological selections. The scheduled monthly instructive lessons are conducted during the work. The antiaccident and fire-fighting trainings are organized not rarely than twice in three months

  19. Current status of NPP generation IV

    International Nuclear Information System (INIS)

    Yohanes Dwi Anggoro; Dharu Dewi; Nurlaila; Arief Tris Yuliyanto

    2013-01-01

    Today development of nuclear technology has reached the stage of research and development of Generation IV nuclear power plants (advanced reactor systems) which is an innovative development from the previous generation of nuclear power plants. There are six types of power generation IV reactors, namely: Very High Temperature Reactor (VHTR), Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), and Super Critical Water-cooled Reactor (SCWR). The purpose of this study is to know the development of Generation IV nuclear power plants that have been done by the thirteen countries that are members of the Gen IV International Forum (GIF). The method used is review study and refers to various studies related to the current status of research and development of generation IV nuclear power. The result of this study showed that the systems and technology on Generation IV nuclear power plants offer significant advances in sustainability, safety and reliability, economics, and proliferation resistance and physical protection. In addition, based on the research and development experience is estimated that: SFR can be used optimally in 2015, VHTR in 2020, while NPP types GFR, LFR, MSR, and SCWR in 2025. Utilization of NPP generation IV said to be optimal if fulfill the goal of NPP generation IV, such as: capable to generate energy sustainability and promote long-term availability of nuclear fuel, minimize nuclear waste and reduce the long term stewardship burden, has an advantage in the field of safety and reliability compared to the previous generation of NPP and VHTR technology have a good prospects in Indonesia. (author)

  20. AGNES - safety reassessment of Paks NPP

    International Nuclear Information System (INIS)

    Gado, J.

    1995-01-01

    The main goal of the AGNES (Advanced General and New Evaluation of Safety) project for the reassessment of the safety of Paks Nuclear Power Plant, Hungary, was to improve the safety culture of the technology at Paks. A report was prepared on the reassessment of the Paks NPP safety. The analysis was divided into four groups: systems analysis, analysis of design basis accidents, severe accident analysis, and level 1 probabilistic safety analysis. Proposed safety enhancement measures are discussed. (N.T.)