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Sample records for 300mwe pwr npp

  1. IPSN expert appraisal programme on the chooz A 300 MWe PWR. Lessons learned by IPSN

    Morlent, O.; Reuchet, J. [CEA Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire, 92 (France)

    2001-07-01

    The closure of Chooz A PWR provided an opportunity to take samples of items that had aged in situ in conditions close to those encountered in PWR in operation over a period of 140.000 hours, which is far longer than the usual time-spans of simulated laboratory tests. 4 topics have been studied: 1) effect of radiation on reactor vessel internals, 2) dissimilar metal joints of reactor coolant system: pressurizer surge line, 3) cast parts of austeno-ferritic steel: hot and cold leg primary valves, and 4) ageing of cables in high temperatures and under irradiation. The examination of the lower internals on some baffle angle bracket and core shroud screws, subjected to varying amounts of irradiation, did not reveal any cracking or corrosion, and confirmed the saturation effect between 4 and 10 dpa for the hardening of 304 austenitic steel in the low temperature range. Expert appraisal of the dissimilar metal joints on the pressurizer surge line confirmed the existence of small fabrication defects due to high temperature cracking. Expert appraisal of the 3 valve body samples from the main section of the coolant system confirmed that -) thermal ageing of the valve body on the hot leg was more advanced than that of the cold leg valve, -) the material of the valve housing on the cold leg which, in theory, was not sensitive to ageing phenomena, exhibited unexpectedly low impact strength values. As for cables, measurements confirmed that their mechanical and electrical properties remained sufficient for them to carry out their functions. (A.C.)

  2. Improved NOx emissions and combustion characteristics for a retrofitted down-fired 300-MWe utility boiler.

    Li, Zhengqi; Ren, Feng; Chen, Zhichao; Liu, Guangkui; Xu, Zhenxing

    2010-05-15

    A new technique combining high boiler efficiency and low-NO(x) emissions was employed in a 300MWe down-fired boiler as an economical means to reduce NO(x) emissions in down-fired boilers burning low-volatile coals. Experiments were conducted on this boiler after the retrofit with measurements taken of gas temperature distributions along the primary air and coal mixture flows and in the furnace, furnace temperatures along the main axis and gas concentrations such as O(2), CO and NO(x) in the near-wall region. Data were compared with those obtained before the retrofit and verified that by applying the combined technique, gas temperature distributions in the furnace become more reasonable. Peak temperatures were lowered from the upper furnace to the lower furnace and flame stability was improved. Despite burning low-volatile coals, NO(x) emissions can be lowered by as much as 50% without increasing the levels of unburnt carbon in fly ash and reducing boiler thermal efficiency.

  3. CFD investigation on the flow and combustion in a 300 MWe tangentially fired pulverized-coal furnace

    Khaldi, Nawel; Chouari, Yoldoss; Mhiri, Hatem; Bournot, Philippe

    2016-09-01

    The characteristics of the flow, combustion and temperature in a 300 MWe tangentially fired pulverized-coal furnace are numerically studied using computational fluid dynamics. The mathematical model is based on a Eulerian description for the continuum phase and a Lagrangian description for coal particles. The combustion reaction scheme was modeled using eddy dissipation concept. The application of a proper turbulence model is mandatory to generate accurate predictions of flow and heat transfer during combustion. The current work presents a comparative study to identify the suitable turbulence model for tangentially fired furnace problem. Three turbulence models including the standard k-ɛ model, the RNG k-ɛ model and the Reynolds Stress model, RSM are examined. The predictions are compared with the published experimental data of Zheng et al. (Proc Combust Inst 29: 811-818, 2002). The RNG k-ɛ model proves to be the most suitable turbulence model, offering a satisfactory prediction of the velocity, temperature and species fields. The detailed results presented in this paper may enhance the understanding of complex flow patterns and combustion processes in tangentially fired pulverized-coal furnaces.

  4. 300 MWe循环流化床锅炉SNCR系统优化设计%The Optimal Design on the SNCR System in a 300 MWe CFB Boiler

    杜鹏飞; 白杨; 李竞岌; 刘青

    2015-01-01

    随着中国最新环保标准的颁布,一部分循环流化床锅炉必须增设烟气脱硝设备以实现氮氧化物( NOx )的达标排放。针对300 MWe等级的大容量循环流化床( CFB)锅炉,从安全保障、技术实现、经济成本等角度详细比较了现有主流脱硝技术———选择性催化还原( SCR)和选择性非催化还原( SNCR)的可用性,并最终选取以尿素溶液作为还原剂的SNCR系统作为推荐方案。通过计算流体动力学( CFD)模拟方法,探究了不同尿素喷嘴布置方式、锅炉负荷等对尿素分布流场、脱硝效率及逃逸率的影响,并建议每部分离器设置10个喷嘴,且竖向均匀布置于分离器进口段内外侧,并给出了该SNCR系统的工艺优化改进建议。%In order to meet the updated emission regulation in China,some CFB boilers have to install the deNOx de-vices. For the 300MWe CFB boiler,the existing SCR and SNCR technologies are compared and the SNCR technology using urea as reductant is recommended in the view of technical,economic and safety concerns. By employing the CFD method,the different urea distributions inside the cyclone with different layout with different nozzle numbers for each cyclone are calculated and compared. The layout of 10 nozzles both inner and outer walls of cyclone inlet has the optimal performance. At the same time,the modification suggestions on process of urea transport,mixing and injection are provided.

  5. Metallography and Microanalysis of Qinshan PhaseⅠ NPP Spent Fuel Rods

    QIAN; Jin; BIAN; Wei; GUO; Li-na; GUO; Yi-fan; CHU; Feng-min; LIANG; Zheng-qiang

    2015-01-01

    Qinshan PhaseⅠNPP is a first domestic commercial PWR and its fuel rods and fuel assembly were designed and manufactured by China.In order to assess the irradiation properties of the fuel rods,8spent fuel rods which were drew out from 3fuel assemblies were transferred to CIAE hot cells for post irradiation examination(PIE)in 2014.The cladding material of the fuel

  6. The Possibility of Building Nuclear Power Plant Free from Severe Accident Risk PWR NPP with advanced all passive safety cooling systems (AAP SCS)%发展无严重事故风险核电站的曙光具有完全非能动安全冷却系统的压水堆核电站

    肖宏才

    2013-01-01

    A complete set of advanced all passive safety cooling systems (AAP SCS) for PWR NPP,actuated by natural force has been put forward in the article.Here the natural force mainly means the fore,which created by change of pressure distribution in the first loop of PWR as a result of operational regime conversion from one to another,including occurrence of accident situation.Correspondent safety cooling system will be actuated naturally and then put it into passive operation after occurring some kind of accident,so accidental situation will be mitigated right after it's occurrence and core residual heat will be naturally moved from the active core to the ultimate heat sink.There is no need to rely on automatic control system,any active equipment and human actions in all working process of the AAP SCS,which can reduce the probability of severe accident to zero,so as to exclude the need of evacuation plan around AAP nuclear power plant and eliminate the public's concern and doubt about nuclear power safety.Implementation of the AAP SCS concept is only based on use of evolutionary measures and state-of-the-art technology.So at present time it can be used for design of new-type third generation PWR nuclear power plant without severe accident risk,and for modernization of existing second generation nuclear power plant.%本文提出了用自然力直接触发启动压水堆核电站一整套完全非能动的停堆安全冷却系统.这里的自然力主要是指一回路运行工况转换时由于其压力分布变化所形成的压差力.在这一系统中,当进行停堆或发生某种一回路事故工况时,相应的安全冷却系统便自然地投入运行,立即缓解事故后果,将事故时一回路释放的能量及堆芯余热非能动地排入最终热阱.在全过程中不依靠自动控制系统、能动设备及任何人为因素的介入,即可确保对堆芯余热无限期的安全冷却能力,完全避免压水堆核电站发生向环境泄漏放射性物

  7. Selection Methodology Approach to Preferable and Alternative Sites for the First NPP Project in Yemen

    Kassim, Moath [Kyunghe Univ., Yongin (Korea, Republic of); Kessel, David S. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-05-15

    The purpose of this paper is to briefly present the methodology and results of the first siting study for the first nuclear power plant (NPP) in Yemen. In this study it has been demonstrated that there are suitable sites for specific unit/units power of 1000 MWt (about 300 MWe) nuclear power plant. To perform the site selection, a systematic selection method was developed. The method uses site-specific data gathered by literature review and expert judgement to identify the most important site selection criteria. A two-step site selection process was used. Candidate sites were chosen that meet a subset of the selection criteria that form the most important system constraints. These candidate sites were then evaluated against the full set of selection criteria using the Analytical Hierarchy Process Method (AHP). Candidate sites underwent a set of more specific siting criteria weighted by expert judgment to select preferable sites and alternatives using AHP method again. Expert Judgment method was used to rank and weight the importance of each criteria, then AHP method used to evaluate and weight the relation between criterion to criterion and between all criteria against the global weight. Then logical decision software was used to rank sites upon their weighting value.

  8. Standard PWR for Italy

    Negroni, A.; Velona, F. (Ente Nazionale per l' Energia Elettrica, Rome (Italy))

    1983-03-01

    A description is given of the general design for the standard PWR which will be used in the seven to eight nuclear power stations provided for in the Italian national energy plan. Special features to meet Italian conditions include double containment and a common foundation mat for the reactor, auxiliary and fuel buildings.

  9. Life management plants at nuclear power plants PWR; Planes de gestion de vida en centrales nucleares PWR

    Esteban, G.

    2014-10-01

    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  10. PWR decontamination feasibility study

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  11. Suomi Npp Products Performance

    Zhou, L.

    2014-12-01

    A suite of Sensor Data Records (SDRs) and Environment Data Records (EDRs) is generated from the Joint Polar Satellite System (JPSS) operational environmental satellite system. The products include atmospheric, ocean, land surface and cryospheric products from the Visible Infrared Imaging Radiometer Suite (VIIRS); atmospheric sounding products from the Cross-track Infrared Sounder (CrIS) and the Advance Technology Microwave Sounder (ATMS); and ozone products from the Ozone Mapping and Profiler Suite (OMPS). These EDRs undergo a rigorous validation process and algorithm updates to achieve a product maturity needed for end user applications. Since the successful launch of Suomi National Polar Partnership (SNPP) satellite in October 2011, significant progresses have been made on calibration and validation of the SNPP data products. By far all products were publicly available and most products were ready for operational evaluation. Most products also are expected to meet requirements and work is underway to reach validated maturity status and fully operational use. Further developments and improvements of the algorithms for J1 have been planned based on the JPSS requirements and lessons learned from SNPP. Sensitivity and impact studies are performed as sensor test data become available. For the majority of data products, no significant changes in sensor input and corresponding sensor degradation are expected. However, the J1 products will undergo the same rigorous calibration and validation process as the S-NPP products once the on-orbit data are available. The schedule for the maturity of the J1 data products however is expected to be accelerated compared to that for S-NPP as lessons learned from the S-NPP mission will be applied to the J1 satellite data. In the presentation, we will provide an overview of the latest SNPP data products' quality status and the plan forward for JPSS-1 algorithm updates.

  12. ANALISIS LAJU DOSIS NEUTRON REAKTOR PLTN PWR 1000 MWe MENGGUNAKAN PROGRAM MCNP

    Amir Hamzah

    2015-03-01

    Full Text Available Dalam rangka menyongsong PLTN pertama di Indonesia, dilakukan kajian dan analisis berbagai aspek teknologi reaktor tersebut. Tujuan dari penelitian ini adalah menentukan laju dosis neutron di luar perisai biologik reaktor PLTN PWR 1000 MWe yang merupakan bagian dari kegiatan besar di atas. Data hasil analisis laju dosis radiasi pada posisi tertentu sangat dibutuhkan untuk menunjukkan tingkat paparan radiasi di posisi tersebut. Analisis laju dosis neutron ditentukan berdasarkan hasil analisis fluks dan spektrum neutron. Analisis fluks dan spektrum neutron di teras reaktor daya PWR 1000 Mwe dilakukan menggunakan program MCNP. Model perhitungan yang dilakukan meliputi 9 zona material yaitu, teras, air, selimut, air, tong, air, bejana tekan, beton dan lapisan udara luar. Penentuan distribusi fluks dan spektrum neutron dilakukan ke arah radial hingga di luar perisai beton dengan akurasi antara 10% hingga 30% dalam tiap kelompok energi yang jumlahnya 1 dan 50 kelompok. Hasil analisis laju dosis neutron di permukaan perisai biologik reaktor PLTN PWR 1000 MWe pada kondisi reaktor beroperasi daya penuh sudah di bawah nilai batas keselamatan. Maka dapat disimpulkan bahwa dari segi paparan radiasi neutron, penggunaan perisai radiasi beton setebal dua meter sudah memenuhi persyaratan keselamatan. Kata kunci: PLTN PWR, fluks neutron, perisai, laju dosis neutron, MCNP.   In order to meet the first nuclear power plant in Indonesia, it has been conducted a study and analysis of various aspects of reactor technology. The purpose of this study was to determine the neutron dose rates at the outside of biological shield of NPP PWR 1000 MWe reactor that is a part of the activities described above. The analysis data of radiation dose rate at a specific position is needed to show the level of radiation exposure in those positions. Analysis neutron dose rate is determined based on the results of the analysis of neutron flux. Analysis of flux and neutron spectrum in

  13. The continued development of the MFM suite and its practical application on a PWR system

    Thunem, Harald P-J; Zhang, Xinxin

    2015-01-01

    This paper reports on the results from the practical application of the Shape Shifter framework on the continued development of a graphical editing suite, the MFM Suite, for MFM and process model design and analysis. The primary use of the MFM Suite is diagnosis and prognosis of anomalies...... in physical processes. One of the Halden Reactor Project’s advanced NPP simulators based on a PWR is used to demonstrate the applicability of the suite in realistic situations. The paper presents a summary and suggests some plans for future research and development....

  14. Physics of hydride fueled PWR

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  15. Development, verification and validation of an FPGA-based core heat removal protection system for a PWR

    Wu, Yichun, E-mail: ycwu@xmu.edu.cn [College of Energy, Xiamen University, Xiamen 361102 (China); Shui, Xuanxuan, E-mail: 807001564@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Cai, Yuanfeng, E-mail: 1056303902@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Zhou, Junyi, E-mail: 1032133755@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Wu, Zhiqiang, E-mail: npic_wu@126.com [State Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Zheng, Jianxiang, E-mail: zwu@xmu.edu.cn [College of Energy, Xiamen University, Xiamen 361102 (China)

    2016-05-15

    Highlights: • An example on life cycle development process and V&V on FPGA-based I&C is presented. • Software standards and guidelines are used in FPGA-based NPP I&C system logic V&V. • Diversified FPGA design and verification languages and tools are utilized. • An NPP operation principle simulator is used to simulate operation scenarios. - Abstract: To reach high confidence and ensure reliability of nuclear FPGA-based safety system, life cycle processes of discipline specification and implementation of design as well as regulations verification and validation (V&V) are needed. A specific example on how to conduct life cycle development process and V&V on FPGA-based core heat removal (CHR) protection system for CPR1000 pressure water reactor (PWR) is presented in this paper. Using the existing standards and guidelines for life cycle development and V&V, a simplified FPGA-based CHR protection system for PWR has been designed, implemented, verified and validated. Diversified verification and simulation languages and tools are used by the independent design team and the V&V team. In the system acceptance testing V&V phase, a CPR1000 NPP operation principle simulator (OPS) model is utilized to simulate normal and abnormal operation scenarios, and provide input data to the under-test FPGA-based CHR protection system and a verified C code CHR function module. The evaluation results are applied to validate the under-test FPGA-based CHR protection system. The OPS model operation outputs also provide reasonable references for the tests. Using an OPS model in the system acceptance testing V&V is cost-effective and high-efficient. A dedicated OPS, as a commercial-off-the-shelf (COTS) item, would contribute as an important tool in the V&V process of NPP I&C systems, including FPGA-based and microprocessor-based systems.

  16. Does climate directly influence NPP globally?

    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.

  17. Suomi NPP Ground System Performance

    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

  18. Decommissioning Study of Oskarshamn NPP

    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.

  19. Decommissioning study of Forsmark NPP

    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.

  20. Integrated ageing management of Atucha NPP

    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)

  1. Forest NPP estimation based on MODIS data under cloudless condition

    CHEN LiangFu; GAO YanHua; LI Li; LIU QinHuo; GU XingFa

    2008-01-01

    Based on light-use efficiency model, an MODIS-derived daily net primary production (NPP) model was developed. In this model, a new model for the fraction of photosynthetically active radiation absorbed by vegetation (FPAR) is developed based on leaf area index (LAI) and albedo parameters, and a photosynthetically active radiation (PAR) is calculated from the combination of Bird's model with aerosol optical thickness and water vapor derived from cloud free MODIS images. These two models are integrated into our predicted NPP model, whose most parameters are retrieved from MODIS data. In order to validate our NPP model, the observed NPP in the Qianyanzhou station and the Changbai Mountains station are used to compare with our predicted NPP, showing that they are in good agreement. The NASA NPP products also have been downloaded and compared with the measurements, which shows that the NASA NPP products underestimated NPP in the Qianyanzhou station but overestimated in the Changbai Mountains station in 2004.

  2. Analysis list: npp-13 [Chip-atlas[Archive

    Full Text Available npp-13 Embryo + ce10 http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/target/npp-13....1.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/target/npp-13.5.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/target/npp...-13.10.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/colo/npp-13.Embryo.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/colo/Embryo.gml ...

  3. Analysis list: npp-3 [Chip-atlas[Archive

    Full Text Available npp-3 Embryo + ce10 http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/target/npp-3.1....tsv http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/target/npp-3.5.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/target/npp...-3.10.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/colo/npp-3.Embryo.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/colo/Embryo.gml ...

  4. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  5. Shielding design for PWR in France

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983.

  6. The integrated PWR; Les REP integres

    Gautier, G.M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs

    2002-07-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  7. Projects of Modifications of design for mitigation of accidents outside the design Bases on nuclear Central PWR Siemens-KWU and Westinghouse; Proyectos de Modificaciones de Sieno para Mitigacion de Accidentes fuera de la Bases de Diseno en Centrales Nucleares PWR Siemens-KWU y Westinghouse

    Dominguez Gonzalez, G.; Cano Rodriguez, L. A.; Arguello Tara, A.

    2014-07-01

    Following the accident at the Japanese Fukushima-Daiichi NPP, the different regulators of nuclear power generation have required numerous reports regarding the evaluation and modification of the capacity of the plants to face accidents with severities beyond that established in their Design Bases. Under this new scenario, with multiple new demands and commitments, EA has carried out the required works for the implementation of strategies to mitigate the consequences of beyond Design Basis accidents for utilities owning Siemens-KWU and Westinghouse PWR nuclear power plants. (Author)

  8. Probes for inspections of heat exchanges installed at nuclear power plants type PWR by eddy current method; Sondas para inspecao de trocadores de calor instalados em usinas nucleares tipo PWR pelo metodo de correntes parasitas

    Silva, Alonso F.O. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Enghenharia Mecanica]. E-mail: kauzz21@yahoo.com; Alencar, Donizete A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mail: daa@cdtn.br

    2007-07-01

    From all non destructive examination methods usable to perform integrity evaluation of critical equipment installed at nuclear power plants (NPP), eddy current test (ET) may be considered the most important one, when examining heat exchangers. For its application, special probes and reference calibration standards are employed. In pressurized water reactor (PWR) NPPs, a particularly critical equipment is the steam generator (SG), a huge heat exchanger that contains thousands of U-bend thin wall tubes. Due to its severe working conditions (pressure and temperature), that component is periodically examined by means of ET. In this paper a revision of the operating fundamentals of the main ET probes, used to perform SG inspections is presented. (author)

  9. Knowledge management in the NPP domain

    Nilsen, Svein; Bisio, Rossella; Ludvigsen, Jan Tore

    2004-03-15

    This report gives an outlook on Knowledge Management (KM) activities within NPP related establishments as of today. There may be less activity in the NPP world as compared to many other industrial sectors. Still there is an awakening within the NPP industry demanding that KM should be attended to at a larger scale. The most notable reason for this is maybe an imminent increase in the number of people going into retirement. The types of establishments involved cover the major kinds such as utilities, research institutes and worldwide nuclear organizations. The report sums up a few of those efforts that are presently being implemented. Moreover the report looks at general advancements within the field of knowledge management. Simply stated the endeavours belong to either one of two classes. The first class emphasize the use of technology to solve knowledge management problems. The second class regard knowledge management as a problem pertaining to human factors and organizational issues. This report maintain that knowledge management initiatives should make due considerations to both perspectives. This report also sums up the Halden Reactor Project short term KM initiative. (Author)

  10. Development of the regulatory guide on the management of aging for the operating NPP

    Shin, Tae Myung; Lee, Jae Kyung; Byeon, Chang Soo; Kim, Hyo Soo [Chungju Univ., Cheongju (Korea, Republic of); Kim, Young Ryul; Eun, Hui Kwang [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-03-15

    this is the first final report, and it includes the analysis of PSR regulation status for the operating NPP of foreign countries and thus established scope and method of Korean PSR. The detailed requirements of aging management, as one of the most important factors, is planned to be dealt in the second final report, The result of study so far can be summarized as below, The necessity and feasibility of domestic PSR application is confirmed through the investigation, comparison and review of PSR implementation and regulation status in foreign countries, And, detailed analysis of vaseline guidelines of IAEA performed to establish a skeleton of desirable safety review guideline for Korean NPP. Our own objectives, scope and strategy of review for PWR are roughly set up form factor by factor analysis of PSR implementation experience in foreign countries and background of IAEA guidelines. The essential elements of review for each PSR safety factors are classified and proposed. For efficient review of proposed safety factors and elements, three different option of framework for PSR guidelines are proposed and compared. Through analysis of strength safety factors and elements, three different options of framework for PSR guidelines are proposed and compared. Through analysis of strength and weakness of the three optional frameworks proposed for Korean PSR guideline, the third one is recommended as the best for both owner and regulator. By maintaining the current framework of SAR at maximum and, at the same time, referring the basic concept of IAEA PSR guide, the detailed contents and its review elements are chosen. The standards on the aging was proposed preliminarily for the application of aging evaluation in PSR.

  11. Length Effect on the Thermal Performance of a Heat Pipe for NPP Decay Heat Removal

    Seo, Joseph; Lee, Jae Young [Handong Global University, Pohang (Korea, Republic of)

    2015-10-15

    After Fukushima accident, importance and necessity of passive safety for nuclear power plant have been emphasized. Due to its passive characteristic, heat pipe is seriously considered as an alternative device of the active safety system for removing decay heat from the reactor core. Among many possible applications of heat pipe in NPP, we considered the application to the control rod. In the situation of SBO(Station Black Out) due to BDBA(Beyond Design Basis Accident) in a PWR, control rods are dropped in to nuclear reactor core automatically. Thus, it is expected that applying heat pipe function to control rod can enhance reactor safety by removing decay heat of fuel assembly. Considering the height of the control rod, L/D of the heat pipe would be larger than 400 if the given diameter is assumed to be similar to the diameter of the control rod. Thus, it may not be the matter for small heat pipes, it is necessary to consider the effects of L/D for the large L/D heat pipes. There for, length effect on the thermal performance of heat pipe for decay heat removal was experimentally investigated in this study. Through this study, the L/D effect on the thermal performance of the large L/D heat pipe for nuclear reactor has been studied.

  12. Study of safety relief valve operation under ATWS conditions. [PWR

    Hutmacher, E.S.; Nesmith, B.J.; Brukiewa, J.B.

    1979-06-25

    A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

  13. Forest NPP estimation based on MODIS data under cloudless condition

    2008-01-01

    Based on light-use efficiency model, an MODIS-derived daily net primary production (NPP) model was developed. In this model, a new model for the fraction of photosynthetically active radiation absorbed by vegetation (FPAR) is developed based on leaf area index (LAI) and albedo parameters, and a pho- tosynthetically active radiation (PAR) is calculated from the combination of Bird’s model with aerosol optical thickness and water vapor derived from cloud free MODIS images. These two models are inte- grated into our predicted NPP model, whose most parameters are retrieved from MODIS data. In order to validate our NPP model, the observed NPP in the Qianyanzhou station and the Changbai Mountains station are used to compare with our predicted NPP, showing that they are in good agreement. The NASA NPP products also have been downloaded and compared with the measurements, which shows that the NASA NPP products underestimated NPP in the Qianyanzhou station but overestimated in the Changbai Mountains station in 2004.

  14. Assessment of Severe Accident Depressurization Valve Activation Strategy for Chinese Improved 1000 MWe PWR

    Ge Shao

    2013-01-01

    Full Text Available To prevent HPME and DCH, SADV is proposed to be added to the pressurizer for Chinese improved 1000 MWe PWR NPP with the reference of EPR design. Rapid depressurization capability is assessed using the mechanical analytical code. Three typical severe accident sequences of TMLB’, SBLOCA, and LOFW are selected. It shows that with activation of the SADV the RCS pressure is low enough to prevent HPME and DCH. Natural circulation at upper RPV and hot leg is considered for the rapid depressurization capacity analysis. The result shows that natural circulation phenomenon results in heat transfer from the core to the pipes in RCS which may cause the creep rupture of pipes in RCS and delays the severe accident progression. Different SADV valve areas are investigated to the influence of depressurization of RCS. Analysis shows that the introduction of SADV with right valve area will delay progression of core degradation to RPV failure. Valve area is to be optimized since smaller SADV area will reduce its effect and too large valve area will lead to excessive loss of water inventory in RCS and makes core degradation progression to RPV failure faster without additional core cooling water sources.

  15. The PWR cores management; La gestion des coeurs REP

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Rippert, D. [CEA Cadarache, Departement d' Etudes des Reacteurs, DER, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    2000-01-25

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  16. Characterization of Factors affecting IASCC of PWR Core Internals

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-09-15

    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  17. A pressure drop model for PWR grids

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  18. Degraded core analysis for the PWR

    Gittus, J.H.

    1987-10-01

    The paper presents an analysis of the probability and consequences of degraded core accidents for the PWR. The article is based on a paper which was presented by the author to the Sizewell-B public inquiry. Degraded core accidents are examined with respect to:- the initiating events, safety plant failure, and processes with a bearing on containment failure. Accident types and frequencies are discussed, as well as the dispersion of radionuclides. Accident risks, i.e. individual and societal risks in degraded core accidents are assessed from:- the amount of radionuclides released, the weather, the population distribution, and the accident frequencies. Uncertainties in the assessment of degraded core accidents are also summarized. (U.K.).

  19. Zebra: An advanced PWR lattice code

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

    2012-07-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  20. N-16 monitors: Almaraz NPP experience

    Adrada, J. [Almaraz NPP, Madrid (Spain)

    1997-02-01

    Almaraz Nuclear Power Plant has installed N-16 monitors - one per steam generator - to control the leakage rate through the steam generator tubes after the application of leak before break (LBB) criteria for the top tube sheet (TTS). After several years of operation with the N-16 monitors, Almaraz NPP experience may be summarized as follows: N-16 monitors are very useful to follow the steam generator leak rate trend and to detect an incipient tube rupture; but they do not provide an exact absolute leak rate value, mainly when there are small leaks. The evolution of the measured N-16 leak rates varies along the fuel cycle, with the same trend for the 3 steam generators. This behaviour is associated with the primary water chemistry evolution along the cycle.

  1. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  2. The Terrestrial NPP Simulation in China since 6ka BP

    HE Yong; DONG Wenjie; JI Jinjun; DAN Li

    2005-01-01

    A better understanding of the long-term global carbon cycle required estimate of the changes in terrestrial carbon storage after the last glacial period. The results of simulation at mid-Holocene (MH) from PMIP (Paleoclimate Modeling Intercomparison Project) and the modern data from CRU (Climate Research Unit,East Anglia University, UK) allow us to use the Atmosphere-Vegetation Interaction Model (AVIM) to simulate the Chinese terrestrial net primary productivity (NPP) at 6ka BP and present time. The change of NPP and total NPP in China from now to mid-Holocene are about 54 g m-2yr-1 and 0.63 Pg yr-1,respectively, mainly due to the build-up of temperate forest and tropical rainforest. Chinese terrestrial NPP variation from MH to now is closely related to the variation in intensity of Asian monsoon, which controlled the climate-vegetation pattern change.

  3. Hydrodynamic model of Fukushima-Daiichi NPP Industrial site flooding

    Vaschenko, V N; Gerasimenko, T V; Vachev, B

    2014-01-01

    While the Fukushima-Daiichi was designed and constructed the maximal tsunami height estimate was about 3 m based on analysis of statistical data including Chile earthquake in 1960. The NPP project industrial site height was 10 m. The further deterministic estimates TPCO-JSCE confirmed the impossibility of the industrial site flooding by a tsunami and therefore confirmed ecological safety of the NPP. However, as a result of beyond design earthquake of 11 March 2011 the tsunami height at the shore near the Fukushima-Daiichi NPP reached 15 m. This led to flooding and severe emergencies having catastrophic environmental consequences. This paper proposes hydrodynamic model of tsunami emerging and traveling based on conservative assumptions. The possibility of a tsunami wave reaching 15 m height at the Fukushima-Daiichi NPP shore was confirmed for deduced hydrodynamic resistance coefficient of 1.8. According to the model developed a possibility of flooding is determined not only by the industrial site height, magni...

  4. Stade NPP. Dismantling of the reactor pool

    Scharf, Daniel; Dziwis, Joachim [E.ON Anlagenservice GmbH Nukleartechnik, Gelsenkirchen (Germany); Kemp, Lutz-Hagen [KKW Stade GmbH und Co. oHG, Stade (Germany)

    2012-11-01

    Within the scope of the 4{sup th} partial decommissioning permission of Stade NPP the activated and contaminated structures of the reactor pool had to be dismantled in order to gain a completely non-radioactive reactor pool area for the subsequent clearance measurement of the reactor building. In order to achieve the aim it was intended to remove the activated pool liner sheets, its activated framework and several contaminated ventilation channels made of stainless steel, the concrete walls of the reactor pool entirely or in parts depending on their activation level, as well as the remaining activated carbon steel structures of the reactor pool bottom. Embedded in the concrete walls there were several highly contaminated excore tubes and the contaminated pool top edge, which were intended to be removed to its full extent. The contract of the Stade NPP initiated reactor pool dismantling project had been awarded to E.ON Anlagenservice GmbH (EAS) and its subsupplier sat. Kerntechnik GmbH for the concrete dismantling works and was performed as follows. In order to minimize the radiation level in the main working area in accordance with the ALARA principle, the liner sheets and middle parts of its framework were removed by means of angle grinders first, as they were the most dose rate relevant parts. As a result the primary average radiation level in the reactor pool (measured in a distance of 500 mm from the walls) was lowered from 40 {mu}Sv/h to less than 2 {mu}Sv/h. After the minimization of the radiation level in the working area the main dismantling step started with the cutting of the reactor pool walls in blocks by means of diamond rope cutters. Once a concrete block was cut out, it was transported into the fuel pool by means of a crane and crane fork, examined radiologically, marked area by area and segmented to debris by means of an electrical excavator with a hydraulic chisel. Afterwards the debris and carbon steel parts were fractioned and packed for further

  5. EPC projects for EPR Flamanville 3 NPP

    Diaz, J.I.; Polo, J.; Aymerich, E.; Cubian, B. [Nuclear Generation Department, Iberdrola Ingenieria y Construccion, Avda. Manoteras 20, 28050 Madrid (Spain)

    2010-07-01

    IBERDROLA Ingenieria y Construccion is carrying out a handful of activities in the EPR Flamanville 3 -FA3 NPP- context since 2007 matching oriented to position the company in the emerging marketplace of new nuclear power plants Generation III+, whose expectation for the next years is highly promising. IBERDROLA Ingenieria y Construccion leads 5 EPC -Engineering, Procurement and Commissioning- projects for FA3 NPP from the Nuclear Island till Sea Water Pumping Station as follows: - Design, procurement. fabrication, installation and testing of 21 shell and tubes heat-exchangers for the nuclear island. 12 out of these 17 HXs are conventional and will be designed according to ASME BPV code Section VIII and have to comply with PED 97/23/CE and ESPN. The remaining 5 HXs are nuclear and will be designed according to ASME BPV code Section III and have also to comply with PED and ESPN. - Design, procurement, fabrication and assembly of 9 demineralizers for different plant systems. Three of these Important To Safety (IPS) equipments have been manufactured according with ASME VIII codes and six of them with EN 13445 code plus additional requirements to comply with PED and final client requirements for nuclear island. - Design, fabrication and installation of qualified travelling water screening filters. The equipments furnished will be two nuclear safety qualified filters and associated equipment (cleaning water system and control system). Additionally some auxiliary devices such as grids, automatic trash rakes and stop gates are included in the contract. - Engineering, procurement, fabrication, erection and commissioning for the condensate treatment plant. This system includes a demineralizer tank, 5 filters, reactive injection mixer, pneumatic and manual valves, piping and instrumentation and control systems. - Engineering, procurement erection and commissioning for the electro-chlorination plant to protect the IPS piping for Condensate Water System for FA3. This system

  6. A PWR Thorium Pin Cell Burnup Benchmark

    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.

  7. Conceptual study on advanced PWR system

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  8. Conceptual study of advanced PWR core design

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  9. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  10. Functional Modelling for Fault Diagnosis and its application for NPP

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    The paper presents functional modelling and its application for diagnosis in nuclear power plants.Functional modelling is defined and it is relevance for coping with the complexity of diagnosis in large scale systems like nuclear plants is explained. The diagnosis task is analyzed....... The use of MFM for reasoning about causes and consequences is explained in detail and demonstrated using the reasoning tool the MFM Suite. MFM applications in nuclear power systems are described by two examples a PWR and a FBRreactor. The PWR example show how MFM can be used to model and reason about...

  11. The advanced main control console for next japanese PWR plants

    Tsuchiya, A. [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Ito, K. [Mitsubishi Heavy Industries, Ltd., Nuclear Energy Systems Engineering Center, Yokohama (Japan); Yokoyama, M. [Mitsubishi Electric Corporation, Energy and Industrial Systems Center, Kobe (Japan)

    2001-07-01

    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  12. Spent fuel management of Jose Cabrera NPP

    Blanco Zurro, J.E.; Garcia Costilla, M. [Area de Generacion - Unidad Nuclear, Gas Natural Fenosa, Avda. de San Luis, 77, 28033 Madrid (Spain); Lavara Sanz, A. [Division Nuclear, SOCOIN, P. del Club Deportivo, 1 - Edificio 5, 28223 Pozuelo de Alarcon, Madrid (Spain); Martinez Abad, J.E. [Departamento de Residuos de Alta Actividad, ENRESA, C/ Emilio Vargas, 7, 28043 Madrid (Spain)

    2010-07-01

    The definitive shutdown of Jose Cabrera Nuclear Power Plant took place on 30. of April 2006. From this moment, cooperation agreements between ENRESA and GAS NATURAL FENOSA were established to reach, among others objectives, its decommissioning, 3 years after the shutdown of the reactor. In order to accomplish the Spanish nuclear regulation, a spent fuel management plan was developed. This plan determined that the fuel assemblies placed in the spent fuel pool would be managed by means of their storage in an interim installation. For this reason, an Independent Spent Fuel Storage Installation (ISFSI) was built at plant site, pioneer in Spain by its characteristics of design. Different administrative authorizations from the point of view of nuclear safety as well as from the environmental were required for ISFSI licensing process. The transference and storage of spent fuel was carried out using the HI-STORM 100Z Dry Storage System, developed by HOLTEC INTERNATIONAL. This system, designed for the spent fuel storage in casks, supports abnormal and very hard accident conditions. The system has three main components: Storage Cask (HI-STORM), Transfer Cask (HI-TRAC) and Multipurpose Canister (MPC). In addition to this, the system has a specific Transport Cask (HI-STAR) for the future transport out of the Plant. More than 30 Design Modifications to the system and plant were implemented to solve structural problems and to include safety and ALARA improvements. The transfer of the spent fuel and its emplacement in the ISFSI began on January 2009 and finished on September of that year allowing starting the decommissioning process, three years and a half after Jose Cabrera NPP shutdown. (authors)

  13. Evaluation of PWR and BWR pin cell benchmark results

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Unit Nuclear Energy, Netherlands Energy Research Foundation ECN, Petten (Netherlands)); Hoogenboorm, J.E.; De Leege, P.F.A. (International Reactor Institute IRI, University of Leiden, Leiden (Netherlands)); Van de Voet, J.; Verhagen, F.C.M. (KEMA NV, Arnhem (Netherlands))

    1992-01-01

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary.

  14. Advanced ion exchange resins for PWR condensate polishing

    Hoffman, B. [Rohm and Haas Co. (United States); Tsuzuki, S. [Rohm and Haas Co. (Japan)

    2002-07-01

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  15. Leak before break application in French PWR plants under operation

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  16. Monte Carlo based radial shield design of typical PWR reactor

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  17. ANALISIS MODEL TERAS 3-DIMENSI UNTUK EVALUASI PARAMETER KRITIKALITAS REAKTOR PWR MAJU KELAS 1000 MW

    Tagor Malem Sembiring

    2015-04-01

    Full Text Available Setelah kejadian Fukushima, penggunaan sistem keselamatan pasif menjadi persyaratan yang penting untuk PLTN. PLTN jenis PWR maju kelas 1000 yang didesain oleh Westinghouse, AP1000, memiliki fitur keselamatan pasif disamping sederhana dan modular. Sebelum memilih suatu PLTN, maka perlu dilakukan suatu evaluasi terhadap parameter desainnya. Salah satu parameter yang penting dalam keselamatan adalah kritikalitas teras. Permasalahan pokok dalam mengevaluasi parameter kritikalitas teras AP1000 tidak adanya data komposisi material SS304 dan H2O di daerah reflektor dan diameter penyerap SS304. Dengan demikian tujuan penelitian ini adalah mendapatkan model teras 3-dimensi AP1000 dan siap diaplikasikan dalam evaluasi parameter kritikalitas teras. Hasil perhitungan menunjukkan bahwa komposisi terbaik SS304 dan H2O di reflektor teras bagian atas dan bawah masing-masing 50 vol%, sedangkan diameter penyerap SS304 adalah 0,960 cm. Evaluasi konsentrasi boron kritis menunjukkan perbedaan yang signifikan dengan nilai desain. Meskipun penyebab utama dari perbedaan ini belum diketahui, akan tetapi dapat dibuktikan bahwa konsentrasi boron kritis sangat sensitif dengan densitas UO2. Untuk reaktivitas padam, reaktor AP1000 memiliki margin subkritikalitas teras yang besar untuk satu siklus operasi. Dengan demikian teras yang diusulkan dapat digunakan sebagai acuan untuk evaluasi parameter teras lainnya atau perangkat analitis lainnya dalam rangka mengevaluasi desain reaktor AP1000. Kata kunci: AP1000, kritikalitas, konsentrasi boron kritis, reaktivitas padam   After the Fukushima accident, the use of passive safety system becomes an important requirement for the nuclear power plant (NPP. The advanced PWR NPP with 1000 MW (electric class, designed by Westinghouse, AP1000, a reactor with the passive safety features as well as simple and modular. Before selecting a nuclear power plant, there should be an evaluation of the design parameter. One important parameter in

  18. Validation of MCNP NPP Activation Simulations for Decommissioning Studies by Analysis of NPP Neutron Activation Foil Measurement Campaigns

    Volmert Ben

    2016-01-01

    Full Text Available In this paper, an overview of the Swiss Nuclear Power Plant (NPP activation methodology is presented and the work towards its validation by in-situ NPP foil irradiation campaigns is outlined. Nuclear Research and consultancy Group (NRG in The Netherlands has been given the task of performing the corresponding neutron metrology. For this purpose, small Aluminium boxes containing a set of circular-shaped neutron activation foils have been prepared. After being irradiated for one complete reactor cycle, the sets have been successfully retrieved, followed by gamma-spectrometric measurements of the individual foils at NRG. Along with the individual activities of the foils, the reaction rates and thermal, intermediate and fast neutron fluence rates at the foil locations have been determined. These determinations include appropriate corrections for gamma self-absorption and neutron self-shielding as well as corresponding measurement uncertainties. The comparison of the NPP Monte Carlo calculations with the results of the foil measurements is done by using an individual generic MCNP model functioning as an interface and allowing the simulation of individual foil activation by predetermined neutron spectra. To summarize, the comparison between calculation and measurement serve as a sound validation of the Swiss NPP activation methodology by demonstrating a satisfying agreement between measurement and calculation. Finally, the validation offers a chance for further improvements of the existing NPP models by ensuing calibration and/or modelling optimizations for key components and structures.

  19. Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident

    Budu Andrei Razvan

    2015-01-01

    Full Text Available European Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA. In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the

  20. Evaluation of PWR and BWR pin cell benchmark results

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  1. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  2. A neutronic study of the cycle PWR-CANDU

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  3. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    Quecedo, M.; Serna, J. J.; Weiner, R. A.; Kersting, P. J.

    1999-05-15

    A statistical procedure has been used to evaluate several alternative descriptions of pressurized water reactor (PWR) cladding corrosion behavior, using an extensive database of Improved (low tin) Zr-4 cladding corrosion measurements from fuel irradiated in commercial PWRs. The in-reactor corrosion enhancement factors considered in the model development are based on a comprehensive review of the current literature for PWR cladding corrosion phenomenology and models. In addition, because prediction of PWR cladding corrosion behavior is very sensitive to the values used for the oxide surface temperatures, several models for the forced convection and sub-cooled nucleate boiling (SNB) coolant heat transfer under PWR conditions have also been evaluated. This evaluation determined that the choice of the forced convection heat transfer has the greatest impact on the ability to fit the data. In addition, the SNB heat transfer model used must account for a continuous transition from forced convection conditions to fully developed SNB conditions. With these choices for the heat transfer models, the evaluation determined that the significant in-reactor corrosion enhancement factors are related to the formation of a hydride rim at the cladding outer diameter, the coolant lithium concentration, and the fast neutron fluence (author) (ml)

  4. Studies of a small PWR for onsite industrial power

    Klepper, O.H.; Smith, W.R.

    1977-04-19

    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application.

  5. Methodology for the LABIHS PWR simulator modernization

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  6. Station Blackout Initiated Event Chronology in LWR/HWR NPP

    Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Since the crisis at Fukushima nuclear power plants, a severe accident progression has been recognized as a very important area for an accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of a severe accident progression among the typical pressurized water reactor (PWR), boiling water reactor (BWR) and pressurized heavy water reactor (PHWR). The OPR 1000-like (ABB-CE type PWR), Peach Bottom-like (BWR/4 RCS with a MARK I Containment), and Wolsong1-like (CANDU6 type) plants are selected as reference plants of typical 1000 MWe PWR, 1140MWe BWR, and 600 MWe PHWR, respectively. The design parameters of these plants are quite different. Some of the major different design features of CANDU6 plant from other light water reactors, in terms of a severe accident, are that the plant adopts a duel primary heat transport system and has an additional amount of cooling water in the calandria vessel (calandria tank, CT) and calandria vault (CV). Another feature is that the CT is always submerged in water because the CV is flooded during normal operation. The containment (reactor building, R/B) failure pressure of the CANDU6 plant is considerably lower than that of the typical PWR or BWR4/MARK-I. The containment vessel free volume of MARK-I is much smaller than that of the PWR or CANDU6 plant. Since there is no steam generator (SG) or passive cooling system, the amount of cooling water inventory in BWR4 is relatively less than other plants. Meanwhile the minimum available time of battery power against station blackout (SBO) accident is different among plant types: six hours for BWR4 and four hours for 1000MWe PWR. Therefore, plant responses against the severe core damage scenarios like Fukushima accident are expected to be much different. By identifying plant response signatures, the appropriate correction actions can be developed as part of severe accident management. A SBO scenario, where all off-site power is lost

  7. Improvement of waste release control in French NPP

    Samson, T.; Lucquin, E.; Dupin, M. [EDF/GDL (France); Florence, D. [EDF/GENV (France); Grisot, M. [EDF/CNPE Saint Laurent (France)

    2002-07-01

    The new waste release control in French NPP is more restrictive than the old one and needs heavy investment to bring plants to compliance with it. The great evolutions are a chemical follow up on more chemicals with a higher measurement frequency and with lower maximum concentrations and a specific measurement of carbon 14. Regarding radioactive releases, a new counting has been settled and activity of carbon 14 release is now measured and no longer calculated. The evolution of the French regulation leads to develop specific procedures and analytical techniques in chemistry and in radiochemistry (UV spectrometric methods, carbon 14 measurements,..) EDF NPP operators have launched a voluntarist process to reduce their releases since the beginning and before the evolution of the regulation. EDF priorities in terms of environment care lead henceforth to implement a global optimisation of the impact for a better control of releases. The new regulation will help EDF to reach its goals because it covers all the aspects in one administrative document: it is seen as a real simplification and a clarification towards public. In addition, this new regulation fits in with international practices which will allow an easier comparison of results between EDF and foreign NPP. These big environmental concerns lead EDF to create a national dedicated laboratory (LAMEN) in charge of developing specific measurement procedures to be implemented either by NPP or by sub-contractor laboratories. (authors)

  8. Nuclear Power Plant Module, NPP-1: Nuclear Power Cost Analysis.

    Whitelaw, Robert L.

    The purpose of the Nuclear Power Plant Modules, NPP-1, is to determine the total cost of electricity from a nuclear power plant in terms of all the components contributing to cost. The plan of analysis is in five parts: (1) general formulation of the cost equation; (2) capital cost and fixed charges thereon; (3) operational cost for labor,…

  9. Observation and modeling of NPP for Pinus elliottii plantation in subtropical China

    MA ZeQing; LIU QiJing; WANG HuiMin; LI XuanRan; ZENG HuiQing; XU WenJia

    2008-01-01

    Based on the stem analysis of 59 individuals of Pinus elliottii in combination with tree biomass models,we calculated annual biomass increment of forest plots at Qianyanzhou Ecological Station,Chinese Academy of Sciences in subtropical China. In addition,canopy layer and community NPP were calculated based on 12 years' litter fall data. NPP of the 21-year-old forest was estimated by using the BIOME BGC model; and both measured NPP and estimated NPP were compared with flux data. Community lation between annual litter fall and annual biomass increment; and the litter fall was 1.19 times the biomass increment of living trees. From 1985 to 2005,average NPP and GPP values based on BGC simulated tree layer NPP values. NPP accounted for 30.2% (25.6%-32.9%) of GPP,while NEP accounted for 57.5% (48.1%-66.5%) of tree-layer NPP and 41.74% (37%-52%) of stand NPP. Soil respiration accounted for 77.0% of measured tree NPP and 55.9% of the measured stand NPP. NEE based on eddy covariance method was 12.97% higher than the observed NEP.

  10. A concept of PWR using plate and shell heat exchangers

    Freire, Luciano Ondir; Andrade, Delvonei Alves de, E-mail: luciano.ondir@gmail.com, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  11. Assessment of PWR plutonium burners for nuclear energy centers

    Frankel, A J; Shapiro, N L

    1976-06-01

    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible.

  12. Control of corrosion product transport in PWR secondary cycles

    Sawochka, S.G.; Pearl, W.L. [NWT Corp., San Josa, CA (United States); Passell, T.O.; Welty, C.S. [Electric Power Research Institute, Palo Alto, CA (United States)

    1992-12-31

    Transport of corrosion products to PWR steam generators by the feedwater leads to sludge buildup on the tubesheets and fouling of tube-to-tube support crevices. In these regions, chemical impurities concentrate and accelerate tubing corrosion. Deposit buildup on the tubes also can lead to power generation limitations and necessitate chemical cleaning. Extensive corrosion product transport data for PWR secondary cycles has been developed employing integrating sampling techniques which facilitate identification of major corrosion product sources and assessments of the effectiveness of various control options. Plant data currently are available for assessing the impact of factors such as pH, pH control additive, materials of construction, blowdown, condensate treatment, and high temperature drains and feedwater filtration.

  13. Evaluation of PWR and BWR pin cell benchmark results

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  14. Study on thermal-hydraulics during a PWR reflood phase

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  15. PWR core stablity aganst xenon-induced spatial power oscillation

    Moon, H.J.; Han, K.I. (Korea Advanced Energy Research Inst., Seoul (Republic of Korea))

    1982-06-01

    Stability of a PWR core against xenon-induced axial power oscillation is studied using one-dimensional xenon transient analysis code, DD1D, that has been developed and verified at KAERI. Analyzed by DD1D utilizing the Kori Unit 1 design and operating data is the sensitivity of axial stability in a PWR core to the changes in core physical parameters including core power level, moderator temperature coefficient, core inlet temperature, doppler power coefficient and core average burnup. Through the sensitivity study the Kori Unit 1 core is found to be stable against axial xenon oscillation at the beginning of cycle 1. But, it becomes less stable as burnup progresses, and unstable at the end of cycle. Such a decrease in stability is mainly due to combined effect of changes in axial power distribution, moderator temperature coefficient and doppler power coefficient as core burnup progresses. It is concluded from the stability analysis of the Kori Unit 1 core that design of a large PWR with high power density and increased dimension can not avoid xenon-induced axial power instabilites to some extents, especially at the end of cycle.

  16. PWR Cross Section Libraries for ORIGEN-ARP

    McGraw, Carolyn [Texas A& M University; Ilas, Germina [ORNL

    2012-01-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.

  17. Actinides transmutation - a comparison of results for PWR benchmark

    Claro, Luiz H. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)], e-mail: luizhenu@ieav.cta.br

    2009-07-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO{sub 2} used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k{infinity} and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  18. Validation of gadolinium burnout using PWR benchmark specification

    Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2014-07-01

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  19. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR.

  20. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  1. Mapping and analysing cropland use intensity from a NPP perspective

    Niedertscheider, Maria; Kastner, Thomas; Fetzel, Tamara; Haberl, Helmut; Kroisleitner, Christine; Plutzar, Christoph; Erb, Karl-Heinz

    2016-01-01

    Meeting expected surges in global biomass demand while protecting pristine ecosystems likely requires intensification of current croplands. Yet many uncertainties relate to the potentials for cropland intensification, mainly because conceptualizing and measuring land use intensity is intricate, particularly at the global scale. We present a spatially explicit analysis of global cropland use intensity, following an ecological energy flow perspective. We analyze (a) changes of net primary production (NPP) from the potential system (i.e. assuming undisturbed vegetation) to croplands around 2000 and relate these changes to (b) inputs of (N) fertilizer and irrigation and (c) to biomass outputs, allowing for a three dimensional focus on intensification. Globally the actual NPP of croplands, expressed as per cent of their potential NPP (NPPact%), amounts to 77%. A mix of socio-economic and natural factors explains the high spatial variation which ranges from 22.6% to 416.0% within the inner 95 percentiles. NPPact% is well below NPPpot in many developing, (Sub-) Tropical regions, while it massively surpasses NPPpot on irrigated drylands and in many industrialized temperate regions. The interrelations of NPP losses (i.e. the difference between NPPact and NPPpot), agricultural inputs and biomass harvest differ substantially between biogeographical regions. Maintaining NPPpot was particularly N-intensive in forest biomes, as compared to cropland in natural grassland biomes. However, much higher levels of biomass harvest occur in forest biomes. We show that fertilization loads correlate with NPPact% linearly, but the relation gets increasingly blurred beyond a level of 125 kgN ha-1. Thus, large potentials exist to improve N-efficiency at the global scale, as only 10% of global croplands are above this level. Reallocating surplus N could substantially reduce NPP losses by up to 80% below current levels and at the same time increase biomass harvest by almost 30%. However, we

  2. The status of nuclear waste from NPP in Romania

    Mauna, T. [Romanian Nuclear Energy Association Council, Asociatia Romana Energia Nucleara AREN, Bucharest (Romania)]. E-mail: tmauna@nuclearelectrica.ro

    2006-07-01

    AREN founded in 1990 is a Romanian NGO focused to sustain its employees or corporate members to develop all kinds of nuclear activities in connection with environmental protection as a scientific organization, having as the first objective activities with respect to Cernavoda NPP. As the only CANDU type reactor equipped Nuclear Power Plant (NPP) in Europe, we pay very much attention to all aspects regarding implementation of this concept in our country and the consequences of this implementation. From July 1996 the first unit in operation supplied into the grid around 40 TWh electric power and around 400 Tcal of thermal power for district heating until September 2004. The second unit is still under construction managed also by the Canadian project team, having a finalization target year of 2007. The temporary LILW, and spent fuel dry storage facilities are also on Cernavoda NPP site inside the safety exclusion area boundary of the first unit. The capacity of temporary LILW warehouse concrete building, practically located into the security plant fence, is around 2,400 m{sup 3}. T he occupied capacity is estimated as 140 m{sup 3} until the end of 2004. The spent fuel dry storage MACSTOR type (a Canadian solution for spent fuel storage) with about 12,000 spent fuel bundles capacity is in operation on Cernavoda NPP site, since May 2003. Nuclearelectrica as the owner implemented all the projects based on the licenses and permits granted by the National Commission for Nuclear Activities Control (CNCAN) for each step: the sitting, construction, commissioning and operation. According to the specific Romanian regulations, every project on the site, like the interim dry storage facility, was also subject to the licensing process by the Environmental and Public Health authorities. The public acceptance has been an important step of the licensing procedure. Cernavoda NPP used different legal procedures for public debate including announcements in local and national newspapers

  3. Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

    Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)

  4. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  5. Robot dispatching Scenario for Accident Condition Monitoring of NPP

    Kim, Jongseog [Central Research Institute of Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2013-05-15

    In March of 2011, unanticipated big size of tsunami attacks Fukushima NPP, this accident results in explosion of containment building. Tokyo electric power of Japan couldn't dispatch a robot for monitoring of containment inside. USA Packbot robot used for desert war in Iraq was supplied to Fukushima NPP for monitoring of high radiation area. Packbot also couldn't reach deep inside of Fukushima NPP due to short length of power cable. Japanese robot 'Queens' also failed to complete a mission due to communication problem between robot and operator. I think major reason of these robot failures is absence of robot dispatching scenario. If there was a scenario and a rehearsal for monitoring during or after accident, these unanticipated obstacles could be overcome. Robot dispatching scenario studied for accident of nuclear power plant was described herein. Study on scenario of robot dispatching is performed. Flying robot is regarded as good choice for accident monitoring. Walking robot with arm equipped is good for emergency valve close. Short time work and shift work by several robots can be a solution for high radiation area. Thin and soft cable with rolling reel can be a good solution for long time work and good communication.

  6. Integrated tool for NPP lifetime management in Spain

    Francia, L. [UNESA, Madrid (Spain); Lopez de Santa Maria, J. [ASCO-Vandellos 2 NPPs l' Hospitalet de l' Infant, Tarragona (Spain); Cardoso, A. [Tecnatom SA, Madrid (Spain)

    2001-07-01

    The project for the Integrated Nuclear Power Plant Lifetime Management System SIGEVI (Sistema Integrado de GEstion de VIda de Centrales Nucleares) was initiated in April 1998 and finalized in December 2000, the main objective of the project being to develop a computer application facilitating the assessment of the condition and lifetime of nuclear power plant components. This constituted the second phase of a further-reaching project on NPP Lifetime Management. During the first phase of this project, carried out between 1992 and 1995, the methodology and strategy for the lifetime management of the Spanish NPP's were developed. Among others, degradation phenomena were assessed and the most adequate methods for their monitoring were defined. The SIGEVI Project has been performed under the management of UNESA (Spanish Electricity Association) and with the collaboration of different engineering firms and research institutes (Tecnatom, Empresarios Agrupados, Ufisa, Initec and IIT), with Vandellos II as the pilot plant. The rest of the Spanish NPP's have also actively participated through the Project Steering Committee. The following sections describe the scope, the structure and the main functionalities of the system SIGEVI. (authors)

  7. Feed water distribution pipe replacement at Loviisa NPP

    Savolainen, S.; Elsing, B. [Imatran Voima Loviisa NPP (Finland)

    1995-12-31

    Imatran Voima Oy operates two WWER-440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of feed water distribution (FWD) pipe were observed in 1989. The FWD-pipe T-connection had suffered from severe erosion corrosion damages. Similar damages have been been found also in other WWER-440 type NPPs. In 1989 the nozzles of the steam generator YB11 were inspected. No signs of the damages or signs of erosion were detected. The first damaged nozzles were found in 1992 in steam generators of both units. In 1992 it was started studying different possibilities to either repair or replace the damaged FWD-pipes. Due to the difficult conditions for repairing the damaged nozzles it was decided to study different FWD-pipe constructions. In 1991 two new feedwater distributors had been implemented at Dukovany NPP designed by Vitckovice company. Additionally OKB Gidropress had presented their design for new collector. In spring 1994 all the six steam generators of Rovno NPP unit 1 were replaced with FWD-pipes designed by OKB Gidropress. After the implementation an experimental program with the new systems was carried out. Due to the successful experiments at Rovno NPP Unit 1 it was decided to implement `Gidropress solution` during 1994 refueling outage into the steam generator YB52 at Loviisa 2. The object of this paper is to discuss the new FWD-pipe and its effects on the plant safety during normal and accident conditions. (orig.).

  8. EPRI PWR Safety and Relief Valve Test Program: test condition justification report

    Hosler, J.

    1982-12-01

    In response to NUREG 0737, Item II.D.1.A requirements, several safety and relief valve designs were tested by EPRI under PWR utility sponsorship. Justification that the inlet fluid conditions under which these valve designs were tested are representative of those expected in participating domestic PWR units during FSAR, Extended High Pressure Injection, and Cold Overpressurization events is presented.

  9. PWR safety and relief valve test program. Valve selection/juftification report. Final report

    1982-12-01

    NUREG 0578 required that full-scale testing be performed on pressurizer safety valves and relief valves representative of those in use or planned for use in PWR plants. To obtain valve performance data for the entire population of PWR plant valves, nine safety valves and ten relief valves were selected as a fully representative set of test valves. Justification that the selected valves represent all PWR plant valves was provided by each safety and relief valve manufacturer. Both the valve selection and justification work was performed as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of the PWR utilities in response to the recommendations of NUREG 0578 and the requirements of the NRC. Results of the Safety and Relief Valve Selection and Justification effort is documented in this report.

  10. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.

    2015-07-01

    During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

  11. Visualized Research on Primary Loop Simulation for PWR Nuclear Power Plant%压水堆核电厂一回路仿真可视化研究

    肖瑶; 巫英伟; 苏光辉; 秋穗正

    2013-01-01

    In this study the main equipments and the primary loop of PWR nuclear power plant (NPP) were analyzed in detail.The model of point neutron dynamics,steam generator model with two-phase drift-flux governing equations,3-zone nonequilibrium pressurizer model and 4-quadrant main pump performance model were established.Based on the above models,a NPP simulation program was developed by using mixed programming with FORTRAN90 and Visual C++.The simulation program is of capability to achieve visualized simulation for the main equipments in primary loop and entire system of PWR nuclear power plant.It provides not only the visualized functions of real-time plotting,zooming,etc.,but also the output of numerical results with standard picture and/or text formatting files.Besides,the program was validated by comparing the calculation results of the program developed by authors and those of RELAP5/MOD3.0.%对压水堆核电厂一回路系统及主要设备进行了详细分析,建立了点堆中子动力学模型、两相漂移流蒸汽发生器模型、三区不平衡稳压器模型和主循环泵四象限特性模型,并以此为基础使用FORTRAN90语言和Visual C++语言通过混合编程的方法开发了核电厂仿真分析程序,实现了对压水堆核电厂一回路主要设备及全系统的可视化仿真计算.软件提供实时绘图、缩放等可视化功能,还提供了数据结果的标准图片格式和标准文本格式输出.通过将程序的计算结果与RELAP5/MOD3.0计算结果进行比较,对程序的可靠性进行了验证.

  12. A study on thimble plug removal for PWR plants

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  13. Integral Test Facility PKL: Experimental PWR Accident Investigation

    2012-01-01

    Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circul...

  14. Estimating probable flaw distributions in PWR steam generator tubes

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)

    1997-02-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses.

  15. Vertical Drop Of 21-Pwr Waste Package On Unyielding Surface

    S. Mastilovic; A. Scheider; S.M. Bennett

    2001-01-29

    The objective of this calculation is to determine the structural response of a 21-PWR (pressurized-water reactor) Waste Package (WP) subjected to the 2-m vertical drop on an unyielding surface at three different temperatures. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities in two different WP components. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only.

  16. Comparing global models of terrestrial net primary productivity (NPP): Global pattern and differentiation by major biomes

    Kicklighter, D.W.; Bondeau, A.; Schloss, A.L.; Kaduk, J.; McGuire, A.D.

    1999-01-01

    Annual and seasonal net primary productivity estimates (NPP) of 15 global models across latitudinal zones and biomes are compared. The models simulated NPP for contemporary climate using common, spatially explicit data sets for climate, soil texture, and normalized difference vegetation index (NDVI). Differences among NPP estimates varied over space and time. The largest differences occur during the summer months in boreal forests (50??to 60??N) and during the dry seasons of tropical evergreen forests. Differences in NPP estimates are related to model assumptions about vegetation structure, model parameterizations, and input data sets.

  17. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Kaercher, S.

    2005-07-01

    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  18. Characterization of neutron field in a NPP workplace.

    Breznik, B; Pochat, J L; Muller, H; Asselineau, B; Pavlin, M

    2007-01-01

    At the Krsko Nuclear Power Plant (NPP), albedo dosimeters are used for personal neutron dosimetry. Spectrometric measurements allow determination of reference dosimetric values of realistic neutron fields to be used for calibration of albedo dosimeters. The Laboratory for Neutron Metrology and Dosimetry from the Institute for Radiological Protection and Nuclear Safety (IRSN) was in charge of characterising neutron fields in the plant at two representative points with high neutron and gamma dose rate. Calibration of the dosimeters in the workplace used to be performed only by a spherical survey meter. Based on the reference dosimetric values, the Plant Dosimetry Laboratory has verified the response of albedo dosimeters.

  19. Status of steam generator tubing integrity at Jaslovske Bohunice NPP

    Cepcek, S. [Nuclear Regulatory Authority of the Slovak Republic, Trnava (Slovakia)

    1997-02-01

    Steam generator represents one of the most important component of nuclear power plants. Especially, loss of tubing integrity of steam generators can lead to the primary coolant leak to secondary circuit and in worse cases to the unit shut down or to the PTS events occurrence. Therefore, to ensure the steam generator tubing integrity and the current knowledge about tube degradation propagation and development is of the highest importance. In this paper the present status of steam generator tubing integrity in operated NPP in Slovak Republic is presented.

  20. Degradation of fastener in reactor internal of PWR

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

    2000-03-01

    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  1. VERA Core Simulator Methodology for PWR Cycle Depletion

    Kochunas, Brendan [University of Michigan; Collins, Benjamin S [ORNL; Jabaay, Daniel [University of Michigan; Kim, Kang Seog [ORNL; Graham, Aaron [University of Michigan; Stimpson, Shane [University of Michigan; Wieselquist, William A [ORNL; Clarno, Kevin T [ORNL; Palmtag, Scott [Core Physics, Inc.; Downar, Thomas [University of Michigan; Gehin, Jess C [ORNL

    2015-01-01

    This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclear reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.

  2. SCOR 1000: an economic and innovative conceptual design PWR

    Gautier, G.M.; Chenaud, M.S. [CEA Cadarache (DEN/DER/SESI), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Reacteurs; Tourniaire, B. [CEA Grenoble (DEN/DTN/SE2T/LPTM), 38 (France)

    2007-07-01

    Within the framework of innovative reactors studies, the Cea proposes the SCOR design (Simple COmpact Reactor) based on most of the advantages of innovative reactors. All main components are integrated in the vessel: the pressurizer, the canned pumps, the control rod mechanics of the driving system (CMD), and the dedicated heat exchangers of the passive heat removal system. The only steam generator is located above the vessel instead of the upper head. This design is featured by its compactness and by a large suppression or simplification of auxiliary systems. The first design with a 600 MWe shows its competitiveness with regard to the large loop-type PWR. To reduce the cost investment by the law sized effect, we examine the possibility of increasing the power of the reactor, while keeping the safety advantages of the medium sized SCOR. The electrical power of the new design is 1000 MWe. SCOR-1000 operates at much lower primary circuit pressure than standard PWRs (93 bars instead of the usual 155 bars), and the power density is lower (80 MW/m3 instead of 100 for the present PWRs). The reactivity is controlled by the CMD and by the burnable poison, without soluble boron. With the same safety advantages of the medium-sized SCOR, the cost reduction of the investment and of cost production could reach 18% with regard to the loop-type PWR. (authors)

  3. PWR reactor vessel in-service-inspection according to RSEM

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul [Intercontrole, 13, rue du Capricorne - SILIC 433, 94583 Rungis - Cedex (France)

    2006-07-01

    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high

  4. Degradation and Failure Characteristics of NPP Containment Protective Coating Systems

    Sindelar, R.L.

    2001-04-10

    Nuclear power plants (NPPs) must ensure that the emergency core cooling system (ECCS) or safety-related containment spray system (CSS) remains capable of performing its design safety function throughout the life of the plant. This requires ensuring that long-term core cooling can be maintained following a postulated loss-of-coolant accident (LOCA). Adequate safety operation can be impaired if the protective coatings which have been applied to the concrete and steel structures within the primary containment fail, producing transportable debris which could then accumulate on BWR ECCS suction strainers or PWR ECCS sump debris screens located within the containment. This document will present the data collected during the investigation of coating specimens from plants.

  5. Suomi NPP VIIRS Reflective Solar Bands Operational Calibration Reprocessing

    Slawomir Blonski

    2015-12-01

    Full Text Available Radiometric calibration coefficients for the VIIRS (Visible Infrared Imaging Radiometer Suite reflective solar bands have been reprocessed from the beginning of the Suomi NPP (National Polar-orbiting Partnership mission until present. An automated calibration procedure, implemented in the NOAA (National Oceanic and Atmospheric Administration JPSS (Joint Polar Satellite System operational data production system, was applied to reprocess onboard solar calibration data and solar diffuser degradation measurements. The latest processing parameters from the operational system were used to include corrected solar vectors, optimized directional dependence of attenuation screens transmittance and solar diffuser reflectance, updated prelaunch calibration coefficients without an offset term, and optimized Robust Holt-Winters filter parameters. The parameters were consistently used to generate a complete set of the radiometric calibration coefficients for the entire duration of the Suomi NPP mission. The reprocessing has demonstrated that the automated calibration procedure can be successfully applied to all solar measurements acquired from the beginning of the mission until the full deployment of the automated procedure in the operational processing system. The reprocessed calibration coefficients can be further used to reprocess VIIRS SDR (Sensor Data Record and other data products. The reprocessing has also demonstrated how the automated calibration procedure can be used during activation of the VIIRS instruments on the future JPSS satellites.

  6. Development of NPP Monitoring and Operation Support Technology

    Lee, Jung Woon; Park, Jae Chang; Lee, Yong Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)

    2008-04-15

    During the first stage (2001.7.1-2004.6.30), we developed general human factors design guidelines VDU-based workstations, PMAS alarm display design guidelines, PMAS SPADES display design guidelines, and the revision of KHNP HFE guidelines (HF-010), which have been applied to domestic NPP designs. We also supported other KNICS projects by performing RPS COM design reviews, development of RPS COM Style Guide, and a review of CEDMCS cabinet operator module display design. We developed the ADIOS prototype, NPP performance analysis systems for YGN No.1, 2 plants and Kori No. 2 plant, alarm cause tracking systems for Kori No. 2 plant and OPR1000, and signal fault detection and diagnosis methods for deaerators and steam generators. During the second stage(2004.7.1-2008.4.30), we supported other KNICS projects by reviewing RPS COM display designs three times, developing ESF-CCS COM style guides and reviewing ESF-CCS COM display design, reviewing CRCS LOM and PCS MTP display designs, and developing requirements for DCS GUI components. We also developed integrated style guide for I and C cabinet operator module display designs. In cooperative research with KOPEC-AE, we developed basic technologies for advanced HSI design including task analysis methods, an information and control requirements database, display design criteria, a HSI prototype with its evaluation, and methods for human factors engineering verification and validation.

  7. Modernization and power upgrading of the Loviisa NPP

    Keskinen, A. [IVO Power Engineering Ltd., Vantaa (Finland)

    1997-12-31

    In 1995, Imatran Voima Oy (IVO) started a project for modernization and power upgrading of the Loviisa NPP. The main objectives of the project are to ensure plant safety, to increase electricity production and to improve the expertise of the IVO staff. The total electricity output of Loviisa 1 and 2 units is planned to be increased by about 100 MW. This will be achieved through renovation of the steam turbines and through gradual increase in the thermal reactor power up to 1,500 MW from the present level of 1,375 MW. The Loviisa NPP Final Safety Analysis Report has been revised to a great extent in connection with the licensing process of the reactor power upgrading. The project also includes certain improvements in the primary and safety systems to ensure plant safety. The total cost estimate of the project is around 200 million FIM. The project implementation started in 1995 and in accordance with the plans in 2000 after several phases the last measures at power plant will be completed. (orig.). 4 refs.

  8. Nonvascular contribution to ecosystem NPP in a subarctic heath during early and late growing season

    Campioli, Matteo; Samson, Roeland; Michelsen, Anders

    2009-01-01

    Bryophytes and lichens abound in many arctic ecosystems and can contribute substantially to the ecosystem net primary production (NPP). Because of their growth seasonality and their potential for growth out of the growing season peak, bryophyte and lichen contribution to NPP may be particularly...

  9. Comparative analysis of NPP changes in global tropical forests from 2001 to 2013

    Yin, S.; Li, X.; Wu, W.

    2017-02-01

    Net primary production (NPP) is the difference between total photosynthesis (gross primary production, GPP) and total plant respiration in an ecosystem. NPP is a key component of the terrestrial carbon cycle and is important in global climate research. Tropical forests, distributed mainly in Central Africa, Central and South America, and Southeast Asia, are among the most important ecosystems on earth. They are very important to analyses of the global carbon budget and to the projection of future climatic changes. In this study, we analyzed and compared the temporal and spatial changes of NPP within the three dominant areas of tropical forest from 2001 to 2013 by using data from the Moderate Resolution Imaging Spectroradiometer (MODIS). We found that Central and South America has the highest annual mean NPP, statistically, while the average NPP shows an increasing trend both in Central and South America and Central Africa but a decreasing trend in Southeast Asia.

  10. Modeling local chemistry in PWR steam generator crevices

    Millett, P.J. [EPRI, Palo Alto, CA (United States)

    1997-02-01

    Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledge of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.

  11. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  12. PWR steam generator chemical cleaning, Phase I. Final report

    Rothstein, S.

    1978-07-01

    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.

  13. PWR and BWR spent fuel assembly gamma spectra measurements

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  14. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    M.Krug, R.Shogan

    2004-09-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  15. Development of NPP personnel training system in Ukraine

    Tarykin, V. [Operation Personnel Training Department, Khmelnitsky NPP, Training Center, Neteshin 30100, Khmelnitsky region (Ukraine)]. E-mail: tarykinv@ukr.net

    2005-07-01

    Modern personnel training and retraining system is a guarantee of NPPs safe reliable operation. Since the time when independence of Ukraine was proclaimed personnel training system was created directly at NPPs. This system is based on the latest legislation framework, developed subject to IAEA recommendations, gained international experience in the field of personnel training in view of increased demands to personnel qualification. Training Centers, formed at each plant, form one of the main components of NPP personnel training. Personnel training at Training Centers is performed in accordance with standard programmes. Simulator training base was created by joint efforts of specialists from the USA, Russia and Ukraine. Establishing manager training system and replacement reserves for National Nuclear Energy Generating Company 'ENERGOATOM' (NNEGC 'ENERGOATOM') managerial personnel, including training programme and training materials development, teacher selection and training, is under way. (author)

  16. MORE: Management of Requirements in NPP Modernisation Projects, final report

    Fredriksen, R.; Katta, V.; Raspotnig, C. (Inst. for energiteknikk (IFE) (Norway)); Valkonen, J. (Technical Research Centre of Finland (VTT) (Finland))

    2008-09-15

    This report documents the work and related activities of the MORE (Management of Requirements in NPP Modernisation Projects) (NKS-R project number NKS-R-2005-47) project. This report also provides a summary of the project activities and deliverables, and discusses possible application areas. The project has aimed at the industrial utilisation of the results from the TACO: (Traceability and Communication of Requirements in Digital I and C Systems Development) (NKS-R project number NKS-R-2002-16, completed June, 2005) project, and practical application of improved approaches and methods for requirements engineering and change management. Finally, the report provides a brief description of the extended industrial network and disseminations of the results in Nordic and NKS related events such as seminars and workshops. (author)

  17. Degradation and Failure Characteristics of NPP Containment Protective Coating Systems

    Sindelar, R.L.

    2001-02-22

    A research program to investigate the performance and potential for debris formation of Service Level I coating systems used in nuclear power plant containment is being performed at the Savannah River Technology Center. The research activities are aligned to address phenomena important to cause coating disbondment as identified by the Industry Coatings Expert Panel. The period of interest for performance covers the time from application of the coating through 40 years of service, followed by a medium-to-large break loss-of-coolant accident scenario, which is a design basis accident (DBA) scenario. The interactive program elements are described in this report and the application of these elements to evaluate the performance of the specific coating system of Phenoline 305 epoxy-phenolic topcoat over Carbozinc 11 primer on a steel substrate. This system is one of the predominant coating systems present on steel substrates in NPP containment.

  18. Near Real Time Processing Chain for Suomi NPP Satellite Data

    Monsorno, Roberto; Cuozzo, Giovanni; Costa, Armin; Mateescu, Gabriel; Ventura, Bartolomeo; Zebisch, Marc

    2014-05-01

    Since 2009, the EURAC satellite receiving station, located at Corno del Renon, in a free obstacle site at 2260 m a.s.l., has been acquiring data from Aqua and Terra NASA satellites equipped with Moderate Resolution Imaging Spectroradiometer (MODIS) sensors. The experience gained with this local ground segmenthas given the opportunity of adapting and modifying the processing chain for MODIS data to the Suomi NPP, the natural successor to Terra and Aqua satellites. The processing chain, initially implemented by mean of a proprietary system supplied by Seaspace and Advanced Computer System, was further developed by EURAC's Institute for Applied Remote Sensing engineers. Several algorithms have been developed using MODIS and Visible Infrared Imaging Radiometer Suite (VIIRS) data to produce Snow Cover, Particulate Matter estimation and Meteo maps. These products are implemented on a common processor structure based on the use of configuration files and a generic processor. Data and products have then automatically delivered to the customers such as the Autonomous Province of Bolzano-Civil Protection office. For the processing phase we defined two goals: i) the adaptation and implementation of the products already available for MODIS (and possibly new ones) to VIIRS, that is one of the sensors onboard Suomi NPP; ii) the use of an open source processing chain in order to process NPP data in Near Real Time, exploiting the knowledge we acquired on parallel computing. In order to achieve the second goal, the S-NPP data received and ingested are sent as input to RT-STPS (Real-time Software Telemetry Processing System) software developed by the NASA Direct Readout Laboratory 1 (DRL) that gives as output RDR files (Raw Data Record) for VIIRS, ATMS (Advanced Technology Micorwave Sounder) and CrIS (Cross-track Infrared Sounder)sensors. RDR are then transferred to a server equipped with CSPP2 (Community Satellite Processing Package) software developed by the University of

  19. Application of Moessbauer spectroscopy on corrosion products of NPP

    Dekan, J., E-mail: julius.dekan@stuba.sk; Lipka, J.; Slugen, V. [Institute of Nuclear and Physical Engineering, Faculty of Electrical Engineering and Information Technology, SUT (Slovakia)

    2013-04-15

    Steam generator (SG) is generally one of the most important components at all nuclear power plants (NPP) with close impact to safe and long-term operation. Material degradation and corrosion/erosion processes are serious risks for long-term reliable operation. Steam generators of four VVER-440 units at nuclear power plants V-1 and V-2 in Jaslovske Bohunice (Slovakia) were gradually changed by new original 'Bohunice' design in period 1994-1998, in order to improve corrosion resistance of SGs. Corrosion processes before and after these design and material changes in Bohunice secondary circuit were studied using Moessbauer spectroscopy during last 25 years. Innovations in the feed water pipeline design as well as material composition improvements were evaluated positively. Moessbauer spectroscopy studies of phase composition of corrosion products were performed on real specimens scrapped from water pipelines or in form of filters deposits. Newest results in our long-term corrosion study confirm good operational experiences and suitable chemical regimes (reduction environment) which results mostly in creation of magnetite (on the level 70 % or higher) and small portions of hematite, goethite or hydrooxides. Regular observation of corrosion/erosion processes is essential for keeping NPP operation on high safety level. The output from performed material analyses influences the optimisation of operating chemical regimes and it can be used in optimisation of regimes at decontamination and passivation of pipelines or secondary circuit components. It can be concluded that a longer passivation time leads more to magnetite fraction in the corrosion products composition.

  20. South Ukraine NPP: Safety improvements through Plant Computer upgrade

    Brenman, O. [Westinghouse Electric Company, 4350 Northern Pike, Monroeville, PA 15146 (United States); Chernyshov, M. A. [Westron, LLC, 1 Acad. Proskura St., Kharkiv 61070 (Ukraine); Denning, R. S. [Battelle, 505 King Ave, Columbus, OH 43201 (United States); Kolesov, S. A. [NAEK Energoatom, 3 Vetrov Str., Kiev, 01032 (Ukraine); Balakan, H. H.; Bilyk, B. I.; Kuznetsov, V. I. [PO South Ukraine NPP, NAEK Energoatom, Mylolayv Region, 55000 (Ukraine); Trosman, G. [US Dept. of Energy, International Nuclear Safety Program, Washington, DC 20585 (United States)

    2006-07-01

    This paper summarizes some results of the Plant Computer upgrade at the Units 2 and 3 of South Ukraine Nuclear Power Plant (NPP). A Plant Computer, which is also called the Computer Information System (CIS), is one of the key safety-related systems at VVER-1000 nuclear plants. The main function of the CIS is information support for the plant operators during normal and emergency operational modes. Before this upgrade, South Ukraine NPP operated out-of-date and obsolete systems. This upgrade project wax founded by the U.S. DOE in the framework of the International Nuclear Safety Program (INSP). The most efficient way to improve the quality and reliability of information provided to the plant operator is to upgrade the Human-System Interface (HSI), which is the Upper Level (UL) CIS. The upgrade of the CIS data-acquisition system (DAS), which is the Lower Level (LL) CIS, would have less effect on the unit safety. Generally speaking, the lifetime of the LL CIS is much higher than one of the UL CIS. Unlike Plant Computers at the Western-designed plants, the functionality of the WER-1000 CISs includes a control function (Centralized Protection Testing) and a number of the plant equipment monitoring functions, for example, Protection and Interlock Monitoring and Turbo-Generator Temperature Monitoring. The new system is consistent with a historical migration of the format by which information is presented to the operator away from the traditional graphic displays, for example, Piping and Instrument Diagrams (P and ID's), toward Integral Data displays. The cognitive approach to information presentation is currently limited by some licensing issues, but is adapted to a greater degree with each new system. The paper provides some lessons learned on the management of the international team. (authors)

  1. Information theory-based approach for modeling the cognitive behavior of NPP operators

    Kim, Jong Hyun; Seong, Poong Hyun [KAIST, Taejon (Korea, Republic of)

    2001-10-01

    An NPP system consists of three important components: the machine system, operators, and MMI. Through the MMI, operators monitor and control the plant system. The cognitive model of NPP operators has become a target of modeling by cognitive engineers due to their work environment: complex, uncertain, and safe critical. We suggested the contextual model for the cognitive behavior of NPP operator and the mathematical fundamentals based on information theory which can quantify the model. The demerit of the methodology using the information theory is that it cannot evaluate the correctness and quality of information. Therefore, the validation through the experiment is needed.

  2. Porcine ecto-nucleotide pyrophosphatase/phosphodiesterase 1 (NPP1/CD203a)

    Petersen, Cathrine Bie; Hillig, Ann-Britt Nygaard; Viuff, Birgitte;

    2007-01-01

    /phosphodiesterase 1 (NPP1/CD203a). The porcine NPP1/CD203a encoding gene was mapped to chromosome 1 using a radiation hybrid panel, and transcription was investigated by RT-PCR analysis of several tissues. The cDNA was cloned and introduced into COS7 cells resulting in expression of functionally active enzyme...... and verification of the specificity of an SWC9 reacting monoclonal antibody. The antibody was used for immunohistochemical examination of various porcine tissues. Most prominent expression of NPP1/CD203a was found in lung macrophages and liver sinusoids....

  3. PEMODELAN DAN ANALISIS SEBARAN RADIONUKLIDA DARI PWR PADA KONDISI ABNORMAL DI TAPAK BOJANEGARA-SERANG

    Sri Kuntjoro

    2015-04-01

    Full Text Available Penambahan pembangkit listrik yang baru khususnya pembangkit listrik tenaga nuklir (PLTN berpotensi memberikan konsekuensi radiologis pada masyarakat dan lingkungan, karena adanya lepasan radioaktif dalam kondisi operasi normal maupun abnormal. Oleh karena itu maka pengelola reaktor nuklir harus bisa menyediakan data dan argumentasi yang kuat untuk menjelaskan tentang keselamatan PLTN terhadap lingkungan. Untuk itu perlu dilakukan analisis kondisi abnormal yang terjadi pada PLTN yang akan memberikan konsekuensi radiologis pada lingkungan. Analisis dilakukan dengan membuat pemodelan simulasi kondisi abnormal yang dipostulasikan pada PLTN tipe PWR 1000 MWe serta simulasi dan pemodelan pola potensi lingkungan sebagai daya dukung tapak terhadap penerimaan konsekuensi radiologis tersebut. Pemodelan fenomena transport radionuklida dari teras reaktor sampai ke luar dari sungkup reaktor dilakukan menggunakan perangkat lunak EMERALD dan pemodelan pola dispersi radioaktivitas ke lingkungan dari reaktor meliputi simulasi kondisi meteorologi, distribusi penduduk, produksi dan konsumsi masyarakat pada kondisi ekstrim di daerah studi, menggunakan perangkat lunak GIS, Arcview, Windrose, dan PC COSYMA. Pemodelan konsekuensi radiologis menggunakan tapak contoh daerah Bojanegara-Kramatwatu Pantai Serang-Banten. Dengan menggunakan data sourceterm, data meteorologi dan data dispersi (sebaran penduduk, produksi pertanian dan ternak dan modeling alur paparan (pathway, dihasilkan model sebaran radionuklida dan penerimaan paparan radiasi di lingkungan tapak Bojanegara-Serang, dengan penerimaan dosis radiasi di bawah batas yang diijinkan badan regulator BAPETEN. Kata kunci : PLTN, radioaktivitas, pola dispersi, keselamatan   Additional of electrical power especially Nuclear Power Plant will give radiological consequences to population and environment due to radioactive release in normal and abnormal condition. In consequence the management of nuclear power plant must

  4. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  5. The kinetics of aerosol particle formation and removal in NPP severe accidents

    Zatevakhin, Mikhail A.; Arefiev, Valentin K.; Semashko, Sergey E.; Dolganov, Rostislav A.

    2016-06-01

    Severe Nuclear Power Plant (NPP) accidents are accompanied by release of a massive amount of energy, radioactive products and hydrogen into the atmosphere of the NPP containment. A valid estimation of consequences of such accidents can only be carried out through the use of the integrated codes comprising a description of the basic processes which determine the consequences. A brief description of a coupled aerosol and thermal-hydraulic code to be used for the calculation of the aerosol kinetics within the NPP containment in case of a severe accident is given. The code comprises a KIN aerosol unit integrated into the KUPOL-M thermal-hydraulic code. Some features of aerosol behavior in severe NPP accidents are briefly described.

  6. VIIRS Climate Raw Data Record (C-RDR) from Suomi NPP, Version 1

    National Oceanic and Atmospheric Administration, Department of Commerce — The Suomi NPP Climate Raw Data Record (C-RDR) developed at the NOAA NCDC is an intermediate product processing level (NOAA Level 1b) between a Raw Data Record (RDR)...

  7. Problems of NPP construction project standardization and operational safety increase in the USA

    Gudkov, Yu.V.

    1980-01-01

    Main causes of slackening paces of the USA nuclear power engineering development are considered using the experience of the USA NPP design, construction and operation. Data on increase in capital investments for the NPP construction and variation of their structure to enhancing waste of capital are given. Problems of improvement of operational safety of water-cooled and water-moderated reactors as well as NPP separate assemblies are considered. Ways for reducing periods of NPP design realization are described taking as an example the USA leading power, construction and reactor construction companies. It is shown that the optimization of existing methods for designing NPPs and nuclear reactor structures, systems for construction management and control as well as complex standartization of equipment, construction methods and operational procedures for NPPs are main factors.

  8. Seismic site evaluation practice and seismic design guide for NPP in Continent of China

    Hu Yuxian [State Seismological Bureau, Beijing, BJ (China). Inst. of Geophysics

    1997-03-01

    Energy resources, seismicity, NPP and related regulations of the Continent of China are briefly introduced in the beginning and two codes related to the seismic design of NPP, one on siting and another on design, are discussed in some detail. The one on siting is an official code of the State Seismological Bureau, which specifies the seismic safety evaluation requirements of various kinds of structures, from the most critic and important structures such as NPP to ordinary buildings, and including also engineering works in big cities. The one on seismic design of NPP is a draft subjected to publication now, which will be an official national code. The first one is somewhat unique but the second one is quite similar to those in the world. (author)

  9. Assessment of NPP VIIRS Ocean Color Data Products: Hope and Risk

    Turpie, Kevin R.; Meister, Gerhard; Eplee, Gene; Barnes, Robert A.; Franz, Bryan; Patt, Frederick S.; Robinson, Wayne d.; McClain, Charles R.

    2010-01-01

    For several years, the NASA/Goddard Space Flight Center (GSFC) NPP VIIRS Ocean Science Team (VOST) provided substantial scientific input to the NPP project regarding the use of Visible Infrared Imaging Radiometer Suite (VIIRS) to create science quality ocean color data products. This work has culminated into an assessment of the NPP project and the VIIRS instrument's capability to produce science quality Ocean Color data products. The VOST concluded that many characteristics were similar to earlier instruments, including SeaWiFS or MODIS Aqua. Though instrument performance and calibration risks do exist, it was concluded that programmatic and algorithm issues dominate concerns. Keywords: NPP, VIIRS, Ocean Color, satellite remote sensing, climate data record.

  10. Observation and modeling of NPP for Pinus elliottii plantation in subtropical China

    2008-01-01

    Based on the stem analysis of 59 individuals of Pinus elliottii in combination with tree biomass models, we calculated annual biomass increment of forest plots at Qianyanzhou Ecological Station, Chinese Academy of Sciences in subtropical China. In addition, canopy layer and community NPP were calcu- lated based on 12 years’ litter fall data. NPP of the 21-year-old forest was estimated by using the BIOME BGC model; and both measured NPP and estimated NPP were compared with flux data. Community biomass was 10574 g·m-2; its distribution patterns in tree layer, shrub layer, herbaceous layer, tree root, herbaceous and shrub roots and fine roots were 7542, 480, 239, 1810, 230, 274 and 239 g·m-2, respectively. From 1999 to 2004, the average annual growth rate and litter fall were 741 g·m-2·a-1 (381.31 gC·m-2·a-1) and 849 g·m?2·a?1 (463 gC·m-2·a-1), respectively. There was a significant corre- lation between annual litter fall and annual biomass increment; and the litter fall was 1.19 times the biomass increment of living trees. From 1985 to 2005, average NPP and GPP values based on BGC modeling were 630.88 (343.31 - 906.42 gC·m-2·a-1) and 1 800 gC·m-2·a-1 (1351.62 - 2318.26 gC·m-2·a-1). Regression analysis showed a linear relationship (R2=0.48) between the measured and simulated tree layer NPP values. NPP accounted for 30.2% (25.6%-32.9%) of GPP, while NEP ac- counted for 57.5% (48.1%-66.5%) of tree-layer NPP and 41.74% (37%-52%) of stand NPP. Soil respi- ration accounted for 77.0% of measured tree NPP and 55.9% of the measured stand NPP. NEE based on eddy covariance method was 12.97% higher than the observed NEP.

  11. Radionuclides contamination of fungi after accident on the Chernobyl NPP

    Zarubina, Nataliia E.; Zarubin, Oleg L. [Institute for Nuclear Research of National Academy of Sciense, 03680, pr-t Nauki, 47, Kiev (Ukraine)

    2014-07-01

    Accumulation of radionuclides by the higher fungi (macromycetes) after the accident on the Chernobyl atomic power plant in 1986 has been studied. Researches were spent in territory of the Chernobyl alienation zone and the Kiev region. Our research has shown that macromycetes accumulate almost all types of radionuclides originating from the accident ({sup 131}I, {sup 140}Ba /{sup 140}La, {sup 103}Ru, {sup 106}Ru, {sup 141}Ce, {sup 144}Ce, {sup 95}Nb, {sup 95}Zr, {sup 137}Cs and {sup 134}Cs). They accumulate the long-living {sup 90}Sr in much smaller (to 3 - 4 orders) quantities than {sup 137}Cs. We have established existence of two stages in accumulation of {sup 137}Cs by higher fungi after the accident on the Chernobyl NPP: the first stage resides in the growth of the concentration, the second - in gradual decrease of levels of specific activity of this radionuclide. Despite reduction of {sup 137}Cs specific activity level, the content of this radionuclide at testing areas of the 5-km zone around the Chernobyl NPP reaches 1,100,000 Bq/kg of fresh weight in 2013. We investigated dynamics of accumulation of Cs-137 in higher fungi of different ecological groups. One of the major factors that influence levels of accumulation of {sup 137}Cs by fungi is their nutritional type (ecological group). Fungi that belong to ecological groups of saprotrophes and xylotrophes accumulate this radionuclide in much smaller quantities than symbio-trophic fungi. As a result of the conducted research it has been established that symbio-trophic fungi store more {sup 137}Cs than any other biological objects in forest ecosystems. Among the symbio-trophic fungi species, species showing the highest level of {sup 137}Cs contamination vary in different periods of time after the deposition. It is connected with variability of quantities of these radio nuclides accessible for absorption at the depth of localization of the main part of mycelium of each species in a soil profile. Soil contamination

  12. Development of NPP Safety Requirements into Kenya's Grid Codes

    Ndirangu, Nguni James; Koo, Chang Choong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand.

  13. Mitsubishi PWR nuclear fuel with advanced design features

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  14. Aqueous Nanofluid as a Two-Phase Coolant for PWR

    Pavel N. Alekseev

    2012-01-01

    Full Text Available Density fluctuations in liquid water consist of two topological kinds of instant molecular clusters. The dense ones have helical hydrogen bonds and the nondense ones are tetrahedral clusters with ice-like hydrogen bonds of water molecules. Helical ordering of protons in the dense water clusters can participate in coherent vibrations. The ramified interface of such incompatible structural elements induces clustering impurities in any aqueous solution. These additives can enhance a heat transfer of water as a two-phase coolant for PWR due to natural forming of nanoparticles with a thermal conductivity higher than water. The aqueous nanofluid as a new condensed matter has a great potential for cooling applications. It is a mixture of liquid water and dispersed phase of extremely fine quasi-solid particles usually less than 50 nm in size with the high thermal conductivity. An alternative approach is the formation of gaseous (oxygen or hydrogen nanoparticles in density fluctuations of water. It is possible to obtain stable nanobubbles that can considerably exceed the molecular solubility of oxygen (hydrogen in water. Such a nanofluid can convert the liquid water in the nonstoichiometric state and change its reduction-oxidation (RedOx potential similarly to adding oxidants (or antioxidants for applying 2D water chemistry to aqueous coolant.

  15. PWR safety/relief valve blowdown analysis experience

    Lee, M.Z.; Chou, L.Y.; Yang, S.H. (Gilbert/Commonwealth Engineers and Consultants, Reading, PA (USA). Speciality Engineering Dept.)

    1982-10-01

    The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively. General approaches for generating forcing functions from thermal fluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the 'acceleration or wave force' method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawbacks are discussed. Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.

  16. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  17. Derived Land Surface Emissivity From Suomi NPP CrIS

    Zhou, Daniel K.; Larar, Allen M.; Liu, Xu

    2012-01-01

    Presented here is the land surface IR spectral emissivity retrieved from the Cross-track Infrared Sounder (CrIS) measurements. The CrIS is aboard the Suomi National Polar-orbiting Partnership (NPP) satellite launched on October 28, 2011. We describe the retrieval algorithm, demonstrate the surface emissivity retrieved with CrIS measurements, and inter-comparison with the Infrared Atmospheric Sounding Interferometer (IASI) emissivity. We also demonstrate that surface emissivity from satellite measurements can be used in assistance of monitoring global surface climate change, as a long-term measurement of IASI and CrIS will be provided by the series of EUMETSAT MetOp and US Joint Polar Satellite System (JPSS) satellites. Monthly mean surface properties are produced using last 5-year IASI measurements. A temporal variation indicates seasonal diversity and El Nino/La Nina effects not only shown on the water but also on the land. Surface spectral emissivity and skin temperature from current and future operational satellites can be utilized as a means of long-term monitoring of the Earth's environment. CrIS spectral emissivity are retrieved and compared with IASI. The difference is small and could be within expected retrieval error; however it is under investigation.

  18. Analysis of Steam Generators Corrosion Products from Slovak NPP Bohunice

    Jarmila Degmová

    2012-01-01

    Full Text Available One of the main goals of the nuclear industry is to increase the nuclear safety and reliability of nuclear power plants (NPPs. As the steam generator (SG is the most corrosion sensitive component of NPPs, it is important to analyze the corrosion process and optimize its construction materials to avoid damages like corrosion cracking. For this purpose two different kinds of SGs and its feed water distributing systems from the NPP Jaslovske Bohunice were studied by nondestructive Mössbauer spectroscopy. The samples were scraped from the surface and analyzed in transmission geometry. Magnetite and hematite were found to be the main components in the corrosion layers of both SGs. Dependant of the material the SG consisted of, and the location in the system where the samples were taken, the ratios between magnetite and hematite and the paramagnetic components were different. The obtained results can be used to improve corrosion safety of the VVER-440 secondary circuit as well as to optimize its water chemistry regime.

  19. 3-D CFD Modeling for Parametric Study in a 300-MWe One-Stage Oxygen-Blown Entrained-Bed Coal Gasifier

    Sang Shin Park

    2015-05-01

    Full Text Available Three-dimensional computational fluid dynamics (CFD modeling of the gasification performance in a one-stage, entrained-bed coal gasifier (Shell Coal Gasification Process (SCGP gasifier was performed, for the first time. The parametric study used various O2/coal and steam/coal ratios, and the modeling used a commercial code, ANSYS FLUENT. CFD modeling was conducted by solving the steady-state Navier–Stokes and energy equations using the Eulerian–Lagrangian method. Gas-phase chemical reactions were solved with the Finite–Rate/Eddy–Dissipation Model. The CFD model was verified with actual operating data of Demkolec demo Integrated Gasification Combined Cycle (IGCC facility in Netherlands that used Drayton coal. For Illinois #6 coal, the CFD model was compared with ASPEN Plus results reported in National Energy Technology Laboratory (NETL. For design coal used in the SCGP gasifier in Korea, carbon conversion efficiency, cold gas efficiency, temperature, and species mole fractions at the gasifier exit were calculated and the results were compared with those obtained by using ASPEN Plus-Kinetic. The optimal O2/coal and steam/coal ratios were 0.7 and 0.05, respectively, for the selected operating conditions.

  20. SCC crack growth rate of cold-worked austenitic stainless steels in PWR primary water conditions

    Guerre, C.; Raquet, O.; Herms, E. [Commissariat a l' Energie Atomique (CEA), DEN/DPC/SCCME/LECA, Gif-sur-Yvette Cedex (France); Marie, S. [Commissariat a l' Energie Atomique (CEA), DEN/DM2S/SEMT/LISN, Gif-sur-Yvette Cedex (France); Le Calvar, M. [Inst. for Radiological Protection and Nuclear Safety (IRSN), DSR/SAMS, Fontenay-aux-Roses Cedex (France)

    2007-07-01

    Stress corrosion cracking (SCC) of stainless steels (SS) is a significant cause of failure in the pressurized water reactors (PWR). Most of the reported case history failures of SS in PWR can be attributed to pollutants (chloride, sulphate) and / or locally oxygenated environments, even to sensitisation of the SS. However, some failures have been attributed to heavy cold work (CW) of SS. In laboratory tests, SCC initiation of cold-worked SS has been obtained using slow strain rate tests (SSRT) in nominal PWR environment. This paper describes constant load and cyclic crack growth rate (CGR) tests on cold-worked SS, on CT specimens. 304L and 316L have been tested with a CW up to 60 %. CW 316L is more prone to cracking than 304L. Over 30 % of CW, 316L is susceptible to crack propagation under constant load. CW is the main controlling parameter for cracking. (author))

  1. Nitrogen Limitation is Reducing the Enhancement of NPP by Elevated CO2 in a Deciduous Forest

    Norby, Richard J [ORNL; Warren, Jeffrey [ORNL; Iversen, Colleen M [ORNL; Medlyn, Belinda [Macquarie University; McMurtrie, Ross [University of New South Wales; Hoffman, Forrest M [ORNL

    2008-01-01

    Accurate model representation of the long-term response of forested ecosystems to elevated atmospheric CO2 concentrations (eCO2) is important for predictions of future concentrations of CO2. For biogeochemical models that predict the response of net primary productivity (NPP) to eCO2, free-air CO2 enrichment (FACE) experiments provide the only source of data for comparison. A synthesis of forest FACE experiments reported a 23% increase in NPP in eCO2, and this result has been used as a model benchmark. Here, we provide new evidence from a FACE experiment in a deciduous forest in Tennessee that N limitation has significantly reduced the stimulation of NPP by eCO2, consistent with predictions from ecosystem and global models that incorporate N feedbacks. The Liquidambar styraciflua stand has been exposed to current ambient atmospheric CO2 or air enriched with CO2 to 550 ppm since 1998. Results from the first 6 years of the experiment indicated that NPP was significantly enhanced by eCO2 and that this was a consistent and sustained response. Now, with 10 years of data, our analysis must be revised. The response of NPP to eCO2 has declined from 24% in 2001-2003 to 9% in 2007. The diminishing response to eCO2 since 2004 coincides with declining NPP in ambient CO2 plots. Productivity of this forest stand is limited by N availability, and the steady decline in forest NPP is closely related to changes in the N economy, as evidenced by declining foliar N concentrations. There is a strong linear relationship between foliar [N] and NPP, and the steeper slope in eCO2 indicates that the NPP response to eCO2 should diminish as foliar N declines. Increased fine-root production and root proliferation deeper in the soil have sustained N uptake, but not to an extent sufficient to benefit aboveground production. The mechanistic basis of the N effect on NPP resides in the photosynthetic machinery. The linear relationships between Jmax and Vcmax with foliar [N] did not change from 1998

  2. Nitrogen Limitation is Reducing the Enhancement of NPP by Elevated CO2 in a Deciduous Forest

    Norby, R. J.; Warren, J. M.; Iversen, C. M.; Medlyn, B. E.; McMurtrie, R. E.; Hoffman, F. M.

    2008-12-01

    Accurate model representation of the long-term response of forested ecosystems to elevated atmospheric CO2 concentrations (eCO2) is important for predictions of future concentrations of CO2. For biogeochemical models that predict the response of net primary productivity (NPP) to eCO2, free-air CO2 enrichment (FACE) experiments provide the only source of data for comparison. A synthesis of forest FACE experiments reported a 23% increase in NPP in eCO2, and this result has been used as a model benchmark. Here, we provide new evidence from a FACE experiment in a deciduous forest in Tennessee that N limitation has significantly reduced the stimulation of NPP by eCO2, consistent with predictions from ecosystem and global models that incorporate N feedbacks. The Liquidambar styraciflua stand has been exposed to current ambient atmospheric CO2 or air enriched with CO2 to 550 ppm since 1998. Results from the first 6 years of the experiment indicated that NPP was significantly enhanced by eCO2 and that this was a consistent and sustained response. Now, with 10 years of data, our analysis must be revised. The response of NPP to eCO2 has declined from 24% in 2001-2003 to 9% in 2007. The diminishing response to eCO2 since 2004 coincides with declining NPP in ambient CO2 plots. Productivity of this forest stand is limited by N availability, and the steady decline in forest NPP is closely related to changes in the N economy, as evidenced by declining foliar N concentrations. There is a strong linear relationship between foliar [N] and NPP, and the steeper slope in eCO2 indicates that the NPP response to eCO2 should diminish as foliar N declines. Increased fine-root production and root proliferation deeper in the soil have sustained N uptake, but not to an extent sufficient to benefit aboveground production. The mechanistic basis of the N effect on NPP resides in the photosynthetic machinery. The linear relationships between Jmax and Vcmax with foliar [N] did not change from 1998

  3. Validation of S-NPP VIIRS Sea Surface Temperature Retrieved from NAVO

    Qianguang Tu

    2015-12-01

    Full Text Available The validation of sea surface temperature (SST retrieved from the new sensor Visible Infrared Imaging Radiometer Suite (VIIRS onboard the Suomi National Polar-Orbiting Partnership (S-NPP Satellite is essential for the interpretation, use, and improvement of the new generation SST product. In this study, the magnitude and characteristics of uncertainties in S-NPP VIIRS SST produced by the Naval Oceanographic Office (NAVO are investigated. The NAVO S-NPP VIIRS SST and eight types of quality-controlled in situ SST from the National Oceanic and Atmospheric Administration in situ Quality Monitor (iQuam are condensed into a Taylor diagram. Considering these comparisons and their spatial coverage, the NAVO S-NPP VIIRS SST is then validated using collocated drifters measured SST via a three-way error analysis which also includes SST derived from Moderate Resolution Imaging Spectro-radiometer (MODIS onboard AQUA. The analysis shows that the NAVO S-NPP VIIRS SST is of high accuracy, which lies between the drifters measured SST and AQUA MODIS SST. The histogram of NAVO S-NPP VIIRS SST root-mean-square error (RMSE shows normality in the range of 0–0.6 °C with a median of ~0.31 °C. Global distribution of NAVO VIIRS SST shows pronounced warm biases up to 0.5 °C in the Southern Hemisphere at high latitudes with respect to the drifters measured SST, while near-zero biases are observed in AQUA MODIS. It means that these biases may be caused by the NAVO S-NPP VIIRS SST retrieval algorithm rather than the nature of the SST. The reasons and correction for this bias need to be further studied.

  4. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  5. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    Xiangjie Liu

    2014-01-01

    Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

  6. Modification of the Decontamination Facility at the Kruemmel NPP - 13451

    Klute, Stefan; Kupke, Peter [Siempelkamp Nukleartechnik GmbH Am Taubenfeld 25/1, 69123 Heidelberg (Germany)

    2013-07-01

    In February 2009, Siempelkamp Nukleartechnik GmbH was awarded the contract for the design, manufacture, delivery and construction of a new Decontamination Facility in the controlled area for Kruemmel NPP. The new decontamination equipment has been installed according to the state of art of Kruemmel NPP. The existing space required the following modification, retrofitting and reconstruction works: - Demounting of the existing installation: to create space for the new facility it was necessary to dismantle the old facility. The concrete walls and ceilings were cut into sizes of no more than 400 kg for ease of handling. This enabled decontamination so largest possible amount could be released for recycling. All steel parts were cut into sizes fitting for iron-barred boxes, respecting the requirement to render the parts decontaminable and releasable. - Reconstructing a decontamination facility: Reconstruction of a decontamination box with separate air lock as access area for the decontamination of components and assemblies was conducted using pressurized air with abrasives (glass beads or steel shots). The walls were equipped with sound protection, the inner walls were welded gap-free to prevent the emergence of interstices and were equipped with changeable wear and tear curtains. Abrasive processing unit positioned underneath the dry blasting box adjacent to the two discharge hoppers. A switch has been installed for the separation of the glass beads and the steel shot. The glass beads are directed into a 200 l drum for the disposal. The steel shot was cleaned using a separator. The cleaned steel shot was routed via transportation devices to the storage container, making it available for further blasting operations. A decontamination box with separate air lock as access area for the decontamination of components and assemblies using high pressure water technology was provided by new construction. Water pressures between 160 bar and 800 bar can be selected. The inner

  7. Analyses of PWR boron dilution consequences with the Arrotta code

    Johanson, E.; Cheng, H.W.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced.

  8. Handling of damaged spent fuel at Ignalina NPP

    Ziehm, Ronny [NUKEM Technologies GmbH (Germany); Bechtel, Sascha [Hoefer und Bechtel GmbH (Germany)

    2012-11-01

    The Ignalina Nuclear Power Plant (INPP) is situated in the north-eastern part of Lithuania close to the borders with Latvia and Belarus and on the shore of Lake Druksiai. It is approximately 120 km from the capital city Vilnius. The power plant has two RMBK type water cooled graphite moderated pressure tube reactors each of design capacity 1500MW(e). The start of operation of the Unit 1 was in 1983 and of the Unit 2 in 1987. In the period 1987 - 1991 (i.e. Soviet period) a small proportion of the existing spent nuclear fuel suffered minor to major damages. In the frame of decommissioning of INPP it is necessary that this damaged fuel is retrieved from the storage pools and stored in an interim spent fuel store. NUKEM Technologies GmbH (Germany) as part of a consortium with GNS mbH (Germany) was awarded the contract for an Interim Spent Fuel Storage Facility (B1- ISFSF). This contract includes the design, procurement, manufacturing, supply and installation of a damaged fuel handling system (DFHS). Objective of this DFHS is the safe handling of spent nuclear fuel with major damages, which result in rupture of the cladding and potential loss of fuel pellets from within the cladding. Typical damages are bent fuel bundle skeletons, broken fuel rods, missing or damaged end plugs, very small gaps between fuel bundles, bent central rods between fuel bundles. The presented concept is designed for Ignalina NPP. However, the design is developed more generally to solve these problems with damaged fuel at other nuclear power plants applying these proven techniques. (orig.)

  9. An overview of Suomi NPP VIIRS calibration maneuvers

    Butler, James J.; Xiong, Xiaoxiong; Barnes, Robert A.; Patt, Frederick S.; Sun, Junqiang; Chiang, Kwofu

    2012-09-01

    The first Visible Infrared Imager Radiometer Suite (VIIRS) instrument was successfully launched on-board the Suomi National Polar-orbiting Partnership (SNPP) spacecraft on October 28, 2011. Suomi NPP VIIRS observations are made in 22 spectral bands, from the visible (VIS) to the long-wave infrared (LWIR), and are used to produce 22 Environmental Data Records (EDRs) with a broad range of scientific applications. The quality of these VIIRS EDRs strongly depends on the quality of its calibrated and geo-located Sensor Date Records (SDRs). Built with a strong heritage to the NASA's EOS MODerate resolution Imaging Spectroradiometer (MODIS) instrument, the VIIRS is calibrated on-orbit using a similar set of on-board calibrators (OBC), including a solar diffuser (SD) and solar diffuser stability monitor (SDSM) system for the reflective solar bands (RSB) and a blackbody (BB) for the thermal emissive bands (TEB). Onorbit maneuvers of the SNPP spacecraft provide additional calibration and characterization data from the VIIRS instrument which cannot be obtained pre-launch and are required to produce the highest quality SDRs. These include multiorbit yaw maneuvers for the characterization of SD and SDSM screen transmission, quasi-monthly roll maneuvers to acquire lunar observations to track sensor degradation in the visible through shortwave infrared, and a driven pitch-over maneuver to acquire multiple scans of deep space to determine TEB response versus scan angle (RVS). This paper provides an overview of these three SNPP calibration maneuvers. Discussions are focused on their potential calibration and science benefits, pre-launch planning activities, and on-orbit scheduling and implementation strategies. Results from calibration maneuvers performed during the Intensive Calibration and Validation (ICV) period for the VIIRS sensor are illustrated. Also presented in this paper are lessons learned regarding the implementation of calibration spacecraft maneuvers on follow

  10. An initial assessment of Suomi NPP VIIRS vegetation index EDR

    Vargas, M.; Miura, T.; Shabanov, N.; Kato, A.

    2013-11-01

    The Suomi National Polar-orbiting Partnership (S-NPP) satellite with Visible/Infrared Imager/Radiometer Suite (VIIRS) onboard was launched in October 2011. VIIRS is the primary instrument for a suite of Environmental Data Records (EDR), including Vegetation Index (VI) EDR, for weather forecasting and climate research. The VIIRS VI EDR operational product consists of the Top of the Atmosphere (TOA) Normalized Difference Vegetation Index (NDVI), the Top of the Canopy (TOC) Enhanced Vegetation Index (EVI), and per-pixel product quality information. In this paper, we report results of our assessment of the early VIIRS VI EDR (beta quality) using Aqua MODIS and NOAA-18 AVHRR/3 as a reference for May 2012 to March 2013. We conducted two types of analyses focused on an assessment of physical (global scale) and radiometric (regional scale) performances of VIIRS VI EDR. Both TOA NDVI and TOC EVI of VIIRS showed spatial and temporal trends consistent with the MODIS counterparts, whereas VIIRS TOA NDVI was systematically higher than that of AVHRR. Performance of the early VIIRS VI EDR was limited by a lack of adequate per-pixel quality information, commission/omission errors of the cloud mask, and uncertainties associated with the surface reflectance retrievals. A number of enhancements to the VI EDR are planned, including: (1) implementation of a TOC EVI back-up algorithm, (2) addition of more detailed quality flags on aerosols, clouds, and snow cover, and (3) implementation of gridding and temporal compositing. A web-based, product quality monitoring tool has been developed and automated product validation protocols are being prototyped.

  11. Variations and trends of terrestrial NPP and its relation to climate change in the 10 CMIP5 models

    Suosuo Li; Shihua Lü; Yuanpu Liu; Yanhong Gao; Yinhuan Ao

    2015-03-01

    Using global terrestrial ecosystem net primary productivity (NPP) data, we validated the simulated multi-model ensemble (MME) NPP, analyzed the spatial distribution of global NPP and explored the relationship between NPP and climate variations in historical scenarios of 10 CMIP5 models. The results show that the global spatial pattern of simulated terrestrial ecosystem NPP, is consistent with IGBP NPP, but the values have some differences and there is a huge uncertainty. Considering global climate change, near surface temperature is the major factor affecting the terrestrial ecosystem, followed by the precipitation. This means terrestrial ecosystem NPP is more closely related to near surface temperature than precipitation. Between 1976 and 2005, NPP shows an obvious increasing temporal trend, indicating the terrestrial ecosystem has had a positive response to climate change. MME NPP has increased 3.647PgC during historical period, which shows an increasing temporal trend of 3.9 gCm−2∙100 yr−2 in the past 150 years, also indicating that the terrestrial ecosystem has shown a positive response to climate change in past 150 years.

  12. Development of Information Datasheets of Nuclear Power Plant (NPP) Equipment using cfiXLM schema

    Lee, Jaiho; Song, Eunhye [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In 2009, EPRI (Electrical Power Research Institute) published a new NPP information handover guide to provide NPP owners and operators with data handover templates in consistent format for effective delivery of information during all stages of the handover process. Another difficult concern for NPP data information management is to exchange the data information among many organizations such as NPP owners, operators, engineering companies, suppliers, and vendors. As a matter of fact, the improperly formatted handover of information sometimes occurs due to the discrepancy of data format (e. g., data description language type). This improper delivery can make negative effects on NPP integrity and safety. Thus, the lack of proper exchange for different data information systems of organizations should be resolved by using an international standard data format. The standard data format can reduce the cost and time for data exchange in each phase for design, procurement, delivery, installation, operation and maintenance of equipment. The AEX(automating equipment information exchange) pilot implementation project team under EPRI advanced nuclear technology (ANT) program has been conducted a research for the use of XML equipment schemas for electronic data exchange(EDE). They applied XML equipment schema for the design, selection, quotation, purchase and mock install of a safety injection centrifugal pump using EDE standard HI(hydraulic institute) 50.7. For data exchange, FIATECH, an industry consortium, has equally developed library of templates and reference data for ISO-15926, which is an international standard capable of reducing data-error and delivery time for exchanging data among different organizations. KHNP as an only owner/operator company has not experienced much difficulty in data interoperability with other organizations, but continued its unremitting exertions to develop a robust system capable of managing data information generated in all the stages of NPP

  13. The development of stochastic process modeling through risk analysis derived from scheduling of NPP project

    Lee, Kwang Ho; Roh, Myung Sub [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-10-15

    There are so many different factors to consider when constructing a nuclear power plant successfully from planning to decommissioning. According to PMBOK, all projects have nine domains from a holistic project management perspective. They are equally important to all projects, however, this study focuses mostly on the processes required to manage timely completion of the project and conduct risk management. The overall objective of this study is to let you know what the risk analysis derived from scheduling of NPP project is, and understand how to implement the stochastic process modeling through risk management. Building the Nuclear Power Plant is required a great deal of time and fundamental knowledge related to all engineering. That means that integrated project scheduling management with so many activities is necessary and very important. Simulation techniques for scheduling of NPP project using Open Plan program, Crystal Ball program, and Minitab program can be useful tools for designing optimal schedule planning. Thus far, Open Plan and Monte Carlo programs have been used to calculate the critical path for scheduling network analysis. And also, Minitab program has been applied to monitor the scheduling risk. This approach to stochastic modeling through risk analysis of project activities is very useful for optimizing the schedules of activities using Critical Path Method and managing the scheduling control of NPP project. This study has shown new approach to optimal scheduling of NPP project, however, this does not consider the characteristic of activities according to the NPP site conditions. Hence, this study needs more research considering those factors.

  14. Temporal variability of the NPP-GPP ratio at seasonal and interannual time scales in a temperate beech forest

    M. Campioli

    2011-09-01

    Full Text Available The allocation of carbon (C taken up by the tree canopy for respiration and production of tree organs with different construction and maintenance costs, life span and decomposition rate, crucially affects the residence time of C in forests and their C cycling rate. The carbon-use efficiency, or ratio between net primary production (NPP and gross primary production (GPP, represents a convenient way to analyse the C allocation at the stand level. In this study, we extend the current knowledge on the NPP-GPP ratio in forests by assessing the temporal variability of the NPP-GPP ratio at interannual (for 8 years and seasonal (for 1 year scales for a young temperate beech stand, reporting dynamics for both leaves and woody organs, in particular stems. NPP was determined with biometric methods/litter traps, whereas the GPP was estimated via the eddy covariance micrometeorological technique.

    The interannual variability of the proportion of C allocated to leaf NPP, wood NPP and leaf plus wood NPP (on average 11% yr−1, 29% yr−1 and 39% yr−1, respectively was significant among years with up to 12% yr−1 variation in NPP-GPP ratio. Studies focusing on the comparison of NPP-GPP ratio among forests and models using fixed allocation schemes should take into account the possibility of such relevant interannual variability. Multiple linear regressions indicated that the NPP-GPP ratio of leaves and wood significantly correlated with environmental conditions. Previous year drought and air temperature explained about half of the NPP-GPP variability of leaves and wood, respectively, whereas the NPP-GPP ratio was not decreased by severe drought, with large NPP-GPP ratio on 2003 due mainly to low GPP. During the period between early May and mid June, the majority of GPP was allocated to leaf and stem NPP, whereas these sinks were of little importance later on. Improved estimation of seasonal GPP and of the

  15. Ageing management of french NPP civil work structures

    Gallitre, E.; Dauffer, D.

    2011-04-01

    This paper presents EDF practice about concrete structure ageing management, from the mechanisms analysis to the formal procedure which allows the French company to increase 900 MWe NPP lifetime until 40 years; it will also introduce its action plan for 60 years lifetime extension. This practice is based on a methodology which identifies every ageing mechanism; both plants feedback and state of the art are screened and conclusions are drawn up into an "ageing analysis data sheet". That leads at first to a collection of 57 data sheets which give the mechanism identification, the components that are concerned and an analysis grid which is designed to assess the safety risk. This analysis screens the reference documents describing the mechanism, the design lifetime hypotheses, the associated regulation or codification, the feedback experiences, the accessibility, the maintenance actions, the repair possibility and so one. This analysis has to lead to a conclusion about the risk taking into account monitoring and maintenance. If the data sheet conclusion is not clear enough, then a more detailed report is launched. The technical document which is needed, is a formal detailed report which summarizes every theoretical knowledge and monitoring data: its objective is to propose a solution for ageing management: this solution can include more inspections or specific research development, or additional maintenance. After a first stage on the 900 MWe units, only two generic ageing management detailed reports have been needed for the civil engineering part: one about reactor building containment, and one about other structures which focuses on concrete inflating reactions. The second stage consists on deriving this generic analysis (ageing mechanism and detailed reports) to every plant where a complete ageing report is required (one report for all equipments and structures of the plant, but specific for each reactor). This ageing management is a continuous process because the

  16. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  17. Operating and mathematical representation of resonances between flow parameters oscillations and structure vibrations of NPP

    Proskuryakov, K.N.; Yang Shan Afshar, E.; Polyakov, N.I. [Nuclear Power Plant Department of Moscow Power Engineering Institute Technical Univ., Moscow (Russian Federation)

    2007-07-01

    The experimental data that have been obtained from the measurements of noise signals in primary circuit of NPP with reactor of WWER-1000 are presented. The causes of resonant interaction between Eigen-Frequencies of Oscillations of the Coolant Pressure (EFOCP) and structure vibrations are discussed. An application-oriented approach to the problem of identification of abnormal phenomena of thermal-hydraulic parameters is proposed. Logarithmic Decrement {delta} is determined. The bigger damping ratio {zeta} provides bigger {delta} and correspondingly smaller values of Q-factor and amplitude X(t)max. All experimental units intended for NPP severe accident investigation must satisfy to the NPP Q-factor criterion of similarity. (authors)

  18. Electrical Grid Conditioning For First NPP Integration, a Systems Engineering Approach Incorporating Quality Function Deployment

    Pwani, Henry; James, J. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-10-15

    Nuclear power plant has a high potential to cause serious harm to environment as evidenced by effects of Fukushima and Chernobyl accidents. A reliable electrical power is required for a NPP to facilitate cooling after a shutdown. Failure of electrical power supply during shutdown increases core damage probability. Research shows that a total of 39% of LOOP related events in US are electrical grid centered. In Korea, 38% and 29% of all events that led to NPP shutdown at Hanul units 3-6 and at Hanbit units 3-6 respectively were electrical related. Electric grids for both operating and new NPPs must therefore be examined and upgraded for reliability improvement in order to enhance NPP safety.

  19. Development of DUPIC fuel cycle technology - Assessment of Wolsong NPP fuel handling system for DUPIC fuel

    Na, Bok Gyun; Nam, Gung Ihn [Korea Power Engineering Company, Taejon (Korea)

    2000-04-01

    The DUPIC fuel loading and discharge path of Wolsong NPP is studied assuming that DUPIC fuel is used at Wolsong NPP. Spent DUPIC fuel discharge path is irrelevant, since it uses the same spent fuel discharge path. Number of factors such as safety, economics of design change, radiation exposure to operators, easy of operation and maintenance, etc, are considered in the evaluation of path. A more detailed analysis of cost estimation of the selected path is also carried out. The study shows that DUPIC fuel loading path following through Spent Fuel Storage Bay and Spent Fuel Discharge Port in reverse direction will minimize the design change and additional equipment and radiation exposure to operators. The estimated total cost of using DUPIC fuel in Wolsong NPP based on price index of year 2000 is around 4.5 billion won. 4 refs., 30 figs., 13 tabs. (Author)

  20. Checking of seismic and tsunami hazard for coastal NPP of Chinese continent after Fukushima nuclear accident

    Chang Xiangdong; Zhou Bengang; Zhao Lianda

    2013-01-01

    A checking on seismic and tsunami hazard for coastal nuclear power plant (NPP) of Chinese continent has been made after Japanese Fukushima nuclear accident caused by earthquake tsunami.The results of the checking are introduced briefly in this paper,including the evaluations of seismic and tsunami hazard in NPP siting period,checking results on seismic and tsunami hazard.Because Chinese coastal area belongs to the continental shelf and far from the boundary of plate collision,the tsunami hazard is not significant for coastal area of Chinese continent.However,the effect from tsunami still can' t be excluded absolutely since calculated result of Manila trench tsunami source although the tsunami wave is lower than water level from storm surge.The research about earthquake tsunami will continue in future.The tsunami warning system and emergency program of NPP will be established based on principle of defense in depth in China.

  1. Assessment of void swelling in austenitic stainless steel PWR core internals.

    Chung, H. M.; Energy Technology

    2006-01-31

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling

  2. Counter-regulatory phosphatases TNAP and NPP1 temporally regulate tooth root cementogenesis

    Laura E Zweifler; Mudita K Patel; Francisco H Nociti Jr; Helen F Wimer; Jose L Milla n; Martha J Somerman; Brian L Foster

    2015-01-01

    Cementum is critical for anchoring the insertion of periodontal ligament fibers to the tooth root. Several aspects of cementogenesis remain unclear, including differences between acellular cementum and cellular cementum, and between cementum and bone. Biomineralization is regulated by the ratio of inorganic phosphate (Pi) to mineral inhibitor pyrophosphate (PPi), where local Pi and PPi concentrations are controlled by phosphatases including tissue-nonspecific alkaline phosphatase (TNAP) and ectonucleotide pyrophosphatase/phosphodiesterase 1 (NPP1). The focus of this study was to define the roles of these phosphatases in cementogenesis. TNAP was associated with earliest cementoblasts near forming acellular and cellular cementum. With loss of TNAP in the Alpl null mouse, acellular cementum was inhibited, while cellular cementum production increased, albeit as hypomineralized cementoid. In contrast, NPP1 was detected in cementoblasts after acellular cementum formation, and at low levels around cellular cementum. Loss of NPP1 in the Enpp1 null mouse increased acellular cementum, with little effect on cellular cementum. Developmental patterns were recapitulated in a mouse model for acellular cementum regeneration, with early TNAP expression and later NPP1 expression. In vitro, cementoblasts expressed Alpl gene/protein early, whereas Enpp1 gene/protein expression was significantly induced only under mineralization conditions. These patterns were confirmed in human teeth, including widespread TNAP, and NPP1 restricted to cementoblasts lining acellular cementum. These studies suggest that early TNAP expression creates a low PPi environment promoting acellular cementum initiation, while later NPP1 expression increases PPi, restricting acellular cementum apposition. Alterations in PPi have little effect on cellular cementum formation, though matrix mineralization is affected.

  3. Comparing the impacts of 2003 and 2010 heatwaves in NPP over Europe

    A. Bastos

    2013-10-01

    Full Text Available In the last decade, Europe was stricken by two outstanding heatwaves, the 2003 event in Western Europe and the recent 2010 episode over Russia. Both extreme events were characterised by record-breaking temperatures, and widespread socio-economic impacts, including significant increments on mortality rates, decreases in crop production and in hydroelectric production. This work aims to assess the influence of both mega-heatwaves on vegetation carbon uptake, using yearly Net Primary Production (NPP and monthly Net Photosynthesis (PsN data derived from satellite imagery obtained from MODIS for the period 2000–2011. In 2010, markedly low productivity was observed over a very large area in Russia, at monthly, seasonal and yearly scales, falling below 50% of average NPP. This decrease in NPP in 2010 was far more intense than the one affecting Western Europe in 2003, which corresponded to 20–30% of the average, and affected a~much larger extent. Total NPP anomalies reached −19 Tg C for the selected regions in France during 2003 and −94 Tg C for western Russia in 2010, which corresponds almost to the magnitude of total NPP anomaly during 2010 for the whole Europe. Overall, the widespread negative PsN anomalies in both regions match the patterns of very high temperature values preceded by a long period of below-average precipitation, leading to strong soil moisture deficits, stressing the role of soil-atmosphere coupling. In the case of 2003 heatwave, results indicate a strong influence of moisture deficits coupled with high temperatures in the response of vegetation, while for the 2010 event very high temperatures appear to be the main driver of very low NPP.

  4. An information theory-based approach to modeling the information processing of NPP operators

    Kim, Jong Hyun; Seong, Poong Hyun [Korea Advanced Institute, Taejon (Korea, Republic of)

    2002-08-01

    This paper proposes a quantitative approach to modeling the information processing of NPP operators. The aim of this work is to derive the amount of the information processed during a certain control task. The focus will be on i) developing a model for information processing of NPP operators and ii) quantifying the model. To resolve the problems of the previous approaches based on the information theory, i.e. the problems of single channel approaches, we primarily develop the information processing model having multiple stages, which contains information flows. Then the uncertainty of the information is quantified using the Conant's model, a kind of information theory.

  5. MODIS GPP/NPP for complex land use area: a case study of comparison between MODIS GPP/NPP and ground-based measurements over Korea

    Shim, C.

    2013-12-01

    The Moderate Resolution Imaging Radiometer (MODIS) Gross Primary Productivity (GPP)/Net Primary Productivity (NPP) has been widely used for the study on global terrestrial ecosystem and carbon cycle. The current MODIS product with ~ 1 km spatial resolution, however, has limitation on the information on local scale environment (Pinus densiflora) agreed well with -0.2% of bias (1.6 gCm-2yr-1). The fairly comparable values of the MODIS here however, cannot assure the quality of the MOD17 over the complex vegetation area of Korea since the ground measurements except the eddy covariance tower flux measurements are highly inconsistent. Therefore, the comprehensive experiments to represents GPP/NPP over diverse vegetation types for a comparable scale of MODIS with a consistent measurement technique are necessary in order to evaluate the MODIS vegetation productivity data over Korea, which contains a large portion of highly heterogeneous vegetation area.

  6. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  7. Vegetation NPP Distribution Based on MODIS Data and CASA Model——A Case Study of Northern Hebei Province

    YUAN Jinguo; NIU Zheng; WANG Chenli

    2006-01-01

    Net Primary Productivity (NPP) is one of the important biophysical variables of vegetation activity, and it plays an important role in studying global carbon cycle, carbon source and sink of ecosystem, and spatial and temporal distribution of CO2. Remote sensing can provide broad view quickly, timely and multi-temporally, which makes it an attractive and powerful tool for studying ecosystem primary productivity, at scales ranging from local to global. This paper aims to use Moderate Resolution Imaging Spectroradiometer (MODIS) data to estimate and analyze spatial and temporal distribution of NPP of the northern Hebei Province in 2001 based on Carnegie-Ames-Stanford Approach (CASA) model.The spatial distribution of Absorbed Photosynthetically Active Radiation (APAR) of vegetation and light use efficiency in three geographical subregions, that is, Bashang Plateau Region, Basin Region in the northwestern Hebei Province and Yanshan Mountainous Region in the Northern Hebei Province were analyzed, and total NPP spatial distribution of the study area in 2001 was discussed. Based on 16-day MODIS Fraction of Photosynthetically Active Radiation absorbed by vegetation (FPAR) product, 16-day composite NPP dynamics were calculated using CASA model; the seasonal dyamics of vegetation NPP in three subregions were also analyzed. Result reveals that the total NPP of the study area in 2001 was 25.1877×106 gC/(m2·a), and NPP in 2001 ranged from 2 to 608 gC/(m2·a), with an average of 337.516 gC/(m2·a). NPP of the study area in 2001 accumulated mainly from May to September (DOY 129-272), high NPP values appeared from June to August (DOY 177-204), and the maximum NPP appeared from late July to mid-August (DOY 209-224).

  8. The Power-weakness Ratios (PWR as a Journal Indicator: Testing the “Tournaments” Metaphor in Citation Impact Studies

    Loet Leydesdorff

    2016-09-01

    Full Text Available Purpose: Ramanujacharyulu developed the Power-weakness Ratio (PWR for scoring tournaments. The PWR algorithm has been advocated (and used for measuring the impact of journals. We show how such a newly proposed indicator can empirically be tested. Design/methodology/approach: PWR values can be found by recursively multiplying the citation matrix by itself until convergence is reached in both the cited and citing dimensions; the quotient of these two values is defined as PWR. We study the effectiveness of PWR using journal ecosystems drawn from the Library and Information Science (LIS set of the Web of Science (83 journals as an example. Pajek is used to compute PWRs for the full set, and Excel for the computation in the case of the two smaller sub-graphs: (1 JASIST+ the seven journals that cite JASIST more than 100 times in 2012; and (2 MIS Quart+ the nine journals citing this journal to the same extent. Findings: A test using the set of 83 journals converged, but did not provide interpretable results. Further decomposition of this set into homogeneous sub-graphs shows that—like most other journal indicators—PWR can perhaps be used within homogeneous sets, but not across citation communities. We conclude that PWR does not work as a journal impact indicator; journal impact, for example, is not a tournament. Research limitations: Journals that are not represented on the “citing” dimension of the matrix—for example, because they no longer appear, but are still registered as “cited” (e.g. ARIST—distort the PWR ranking because of zeros or very low values in the denominator. Practical implications: The association of “cited” with “power” and “citing” with “weakness” can be considered as a metaphor. In our opinion, referencing is an actor category and can be Metaphor in Citation Impact Studies in terms of behavior, whereas “citedness” is a property of a document with an expected dynamics very different from that of

  9. Radiative heat transfer modelling in a PWR severe accident sequence

    Magali Zabiego; Florian Fichot [Institut de Radioprotection et de Surete Nucleaire - BP 3 - 13115 Saint-paul-Lez-Durance (France); Pablo Rubiolo [Westinghouse Science and Technology - 1344 Beulah Road - Pittsburgh - PA 15235 (United States)

    2005-07-01

    a debris bed. In particular, an expression of the conductivity was established in cells in which remaining cylinders and debris particles coexist. In the present document, after a recall of the main lines of the modelling, an application to a reactor sequence is proposed. A severe accident transient with core degradation is simulated. The radiative transfer model is shown to behave properly and to smoothly calculate the transitions between the successive core configurations. A comparison with the more classical Hottel method shows that the present model gives a better prediction of the degradation progression owing to a more accurate estimate of the radial heat transfers. References: [1] M. Zabiego et al., ICARE/CATHARE V1: application to a PWR 900 MWe severe accident sequence, SARJ, Tokyo, 1999; [2] M. Zabiego, F. Fichot, P. Rubiolo Transfert radiatif lors d'une sequence accidentelle dans un coeur de Reacteur a Eau sous Pression, Congres Francais de Thermique, SFT 2004, Presqu'ile de Giens, 25-28 mai 2004. (authors)

  10. [Specific Features of Scots Pine Seeds Formation in the Remote Period after the Chernobyl NPP Accident].

    Geras'kin, S A; Vasiliev, D V; Kuzmenkov, A G

    2015-01-01

    The results of long-term (2007-2011) observations on the quality of seed progeny in Scots pine populations inhabiting the sites within the Bryansk region contaminated as a result of the Chernobyl NPP accident are presented. Formed under the chronic exposure seeds are characterized by a high interannual variability, which is largely determined by weather conditions.

  11. Filtration Algorithms of Untrustworthy Analogous Information in APCS at TPP and NPP

    V. Nazarov

    2012-01-01

    Full Text Available The paper considers filtration algorithms of untrustworthy analogous information in APCS at TTP and NPP that make it possible to identify credibility of information transmitted through communication channels in the form of signals and which are continuously changeable in the regime of real time.

  12. Forward alliance. AREVA's initiative for NPP's LTO projects

    Bergholz, Steffen; Heinz, Benedikt; Rudolph, Juergen [AREVA GmbH, Erlangen (Germany); Quist, Perry [EPZ - Elektriciteits-Productiemaatschappij Zuid-Nederland N.V., Borssele (Netherlands)

    2014-08-15

    Nowadays, the abbreviation LTO (Long Term Operation) is well known in the NPP's world. No other kind of projects can ensure the actual economic and environmental questions of the world of energy production with this high efficiency. Two-thirds of the actual international NPP fleet in operation has reached an age of more than 25 years. The former NPP's design life of normally 40 years has changed to 60 years for new plants. Furthermore, the existing fleet is nominally prepared for an operational period of more the 40 years. AREVA had launched the Forward Alliance initiative which offers different activities to ensure a support of customer's LTO projects. This article will give an overview on the Forward Alliance initiative followed by a technical example how the AREVA Fatigue Concept (AFC) can fulfill the fatigue monitoring topic within potential LTO projects. The successfully realized LTO project for NPP Borssele (Netherlands) shows as a customer example how the Forward Alliance blocks can work.

  13. Post-Launch Calibration Support for VIIRS Onboard NASA NPP Spacecraft

    Xiong, Xiaoxion; Chiang, Kwo-Fu; McIntire, Jeffrey; Schwaller, Matthew; Butler, James

    2011-01-01

    The NPP Instrument Calibration Support Element (NICSE) is one of the elements within the NASA NPP Science Data Segment (SDS). The primary responsibility of NICSE is to independently monitor and evaluate on-orbit radiometric and geometric performance of the Visible Infrared Imaging Radiometer Suite (VIIRS) instrument and to validate its Sensor Data Record (SDR) [1]. The NICSE interacts and works closely with other SDS Product Evaluation and Analysis Tools Elements (PEATE) and the NPP Science Team (ST) and supports their on-orbit data product calibration and validation efforts. The NICSE also works closely with the NPP Instrument Calibration Support Team (NICST) during sensor pre-launch testing in ambient and thermal vacuum environment [2]. This paper provides an overview of NICSE VIIRS sensor post-launch calibration support with a focus on the use of sensor on-board calibrators (OBC) for the radiometric calibration and characterization. It presents the current status of NICSE post-launch radiometric calibration tool development effort based on its design requirements

  14. Replacement of battery in Asco NPP Chargers; Sustitucion de cargadores de baterias en C. N. Asco

    Montero Lansanc, J.

    2013-07-01

    The purpose of this paper is to present the project to replace battery chargers at NPP Asco. It describes the reasons for the replacement, the project approach, the development to date and current status of the project, the economics, and some lessons learned during the process.

  15. Activity level of gross α and gross β in airborne aerosol samples around the Qinshan NPP

    CHEN Bin; YE Jida; CHEN Qianyuan; WU Xiaofei; SONG Weili; WANG Hongfeng

    2007-01-01

    The monitoring results of gross α and gross β activity from 2001 to 2005 for environmental airborne aerosol samples around the Qinshan NPP base are presented in this paper. A total of 170 aerosol samples were collected from monitoring sites of Caichenmen village, Qinlian village, Xiajiawan village and Yangliucun village around the Qinshan NPP base. The measured specific activity of gross α and gross β are in the range of 0.02 ~ 0.38 mBq/m3 and 0.10 ~ 1.81 mBq/m3, respectively, with an average of 0.11 mBq/m3 and 0.45mBq/m3, respectively. They are lower than the average of 0.15 mBq/m3 and 0.52mBq/m3, of reference site at Hangzhou City. It is indicated that the specific activity of gross α and gross β for environmental aerosol samples around the Qinshan NPP base had not been increased in normal operating conditions of the NPP.

  16. Local network deployed around the Kozloduy NPP - a useful tool for seismological monitoring

    Solakov, Dimcho; Simeonova, Stela; Dimitrova, Liliya; Slavcheva, Krasimira; Raykova, Plamena; Popova, Maria; Georgiev, Ivan

    2015-04-01

    Radiation risks may transcend national borders, and international cooperation serves to promote and enhance safety globally by exchanging experience and by improving capabilities to control hazards, to prevent accidents, to respond to emergencies and to mitigate any harmful consequences. International safety standards provide support for states in meeting their obligations under general principles of international law, such as those relating to environmental protection. Seismic safety is a key element of NPP safe operation. Safety and security measures have in common the aim of protecting human life and health and the environment. The Kozloduy NPP site is located in the stable part of the Moesian platform (area of about 50000 km2). From seismological point of view the Moesian platform is the most quite area on the territory of Bulgaria. There are neither historical nor instrumental earthquakes with M>4.5 occurred within the platform. The near region (area with radial extent of 30 km) of the NPP site is characterized with very low seismic activity. The strongest recorded quake is the 1987 earthquake МS=3.6, localized 22 km northwest of the Kozloduy NPP site on the territory of Romania. In line with international practice, the geological, geophysical and seismological characteristics of the region around the site have been investigated for the purpose of evaluating the seismic hazards at the NPP site. A local network (LSN) of sensitive seismographs having a recording capability for micro-earthquakes have been installed around Kozloduy NPP and operated since 1997. The operation and data processing, data interpretation, and reporting of the local micro-earthquake network are linked to the national seismic network (NOTSSI). A real-time data transfer from stations to National Data Center (in Sofia) was implemented using the VPN and MAN networks of the Bulgarian Telecommunication. Real-time and interactive data processing are performed by the Seismic Network Data

  17. Aging Behavior Study and Irradiation Damage Simulation of RPV in NPP

    LIN; Yun; TONG; Zhen-feng; HE; Xin-fu; ZHANG; Chang-yi; NING; Guang-sheng; YANG; Wen

    2015-01-01

    Reactor pressure vessel(RPV)is the critical un-changeable component of the PWR during its service lifetime,which determines the lifetime of the nuclear power plant.And it contains the reactor core,reactor internals and primary coolant circuit.It services in an extremely condition with high temperature,high pressure and neutron irradiation.The property decline of the

  18. Effect of dietary NPP level and phytase supplementation on the laying performance over one year period

    Annamaria Tischler

    2015-06-01

    Full Text Available Our trial was aimed to study the effect of different dietary non-phytin phosphorus (NPP levels with and without phytase enzyme supplementation on laying performance and eggshell quality of Tetra SL-LL in the last 25 weeks of the long-term (17 months egg production. A total of 69 Tetra SL-LL layers were allocated into 3 dietary treatments. Two diets with different levels of NPP (2.45 or 2.15 g/kg, HP and LP, respectively were formulated, and 0 or 300 FTU/kg phytase enzyme was added to low NPP feed (LP and LP+E, respectively. Dietary Ca was uniformly adjusted (38.2 g/kg to feed in each treatment. In the course of the trial, intensity of egg production (%, egg weight (g/egg, number of the broken eggs and feed intake (g/d/bird were recorded. Every 2 weeks 20 eggs per treatment were broken to determine the shell strength and thickness. Our results show that low NPP diet had detrimental effect on the intensity of egg production (P<0.05 and phytase added to the LP diet resulted the lowest number of broken eggs (P<0.05. In conclusion, NPP content of the layer diet can be reduced from 2.45 to 2.15 g/kg in the last 25 weeks of the elongated laying term (12-17 month of laying, if supplemented with 300 FTU/kg phytase enzyme without compromising the egg production, and in the same time it can improve eggshell quality and reduce the number of broken eggs.

  19. Prioritization of Delay Factors for NPP Construction Risk in International Project by Using AHP Methodology

    Hossen, Muhammed Mufazzal; Kang, Sunkoo; Kim, Jonghyun [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    It is crucial for the nuclear power plant project decision makers and management personnel to identify the actual factors of construction delay and their ranking in order to take preventive actions. NPP project is complex in nature and the construction phase is one of the most key phase which is subject to many factors result from many sources. From experience, nuclear projects have faced challenges similar to other complex mega projects with additional nuclear specific issues and life time cost of nuclear reactor is concentrated upfront as capital cost, and therefore delays in construction may become intolerable in terms of both lost revenues and interest on the capital. Budget over-runs and delays on next generation new build nuclear projects in recent years clearly demonstrate that the nuclear industry continues to repeat its failed management and project control processes of the past. Similar to major infra-structure projects, actual completion times can vary substantially from initial estimates but this uncertainty is too crucial to the nuclear industry due to high levels of capital at risk, for every year a project is delayed the levelized cost of electricity increases by approximately 8-10%. causes of delay, to develop a generalized AHP model for delay factors, and to prioritize the risk in different factors in various levels of construction phase in international turnkey NPP project. This paper describes and prioritizes Nuclear Power Plant (NPP) construction schedule delay factor for turnkey international project. This study also determines the different party's importance in percentage behind the construction schedule delay of NPP which constitutes main contractor (28.4%), regulatory authority (27.3%), financial and country factor (23.5%), and utility (20.8%). Decision makers of nuclear industry can understand the significance of different factors on NPP construction phase and they can apply risk informed decision making to avoid unexpected

  20. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.

    1989-01-01

    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  1. Eddy current NDT: a suitable tool to measure oxide layer thickness in PWR fuel rods

    Alencar, Donizete A.; Silva Junior, Silverio F. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail: daa@cdtn.br, e-mail: silvasf@cdtn.br; Vieira, Andre L.P.S. [Industrias Nucleares do Brasil (INB S.A.), Resende, RJ (Brazil). Fabrica de Combustivel Nuclear], e-mail: andre@inb.gov.br; Soares, Adolpho [Technotest Consultoria e Acessoria Ltda., Belo Horizonte, MG (Brazil)], e-mail: adolpho@technotest.com.br

    2009-07-01

    Eddy current is a nondestructive test (NDT) widely used in industry to support integrity analysis of components and equipment. In the nuclear area it is frequently applied to inspect tubes installed in tube exchangers, such as steam generators and condensers in PWR plants, as well as turbine blades. Adequately assisted by means of robotic devices, that inspection method has been pointed as a suitable tool to perform accurate oxide layer thickness measurements in PWR fuel rods. This paper shows some theoretical aspects and physical operating principles of the inspection method, as well as test probes construction details, and the calibration reference standards fabrication processes. Furthermore, some data, experimentally obtained at INB laboratories and other technical information obtained from TECNATOM S.A. are presented, showing the accuracy and efficacy of such NDT method. (author)

  2. EPRI PWR Safety and Relief Value Test Program: safety and relief valve test report

    1982-12-01

    A safety and relief valve test program was conducted by EPRI for a group of participating PWR utilities to respond to the USNRC recommendations documented in NUREG 0578 Section 2.1.2, and as clarified in NUREG 0737 Item II.D.1.A. Seventeen safety and relief valves representative of those utilized in or planned for use in participating domestic PWR's were tested under the full range of selected test conditions. This report contains a listing of the selected test valves and the corresponding as tested test matrices, valve performance data and principal observations for the tested safety and relief valves. The information contained in this report may be used by the participating utilities in developing their response to the above mentioned USNRC recommendations.

  3. DOMINO: A fast 3D cartesian discrete ordinates solver for reference PWR simulations and SPN validation

    Courau, T.; Moustafa, S.; Plagne, L.; Poncot, A. [EDF R and D, 1, Av du General de Gaulle, F92141 Clamart cedex (France)

    2013-07-01

    As part of its activity, EDF R and D is developing a new nuclear core simulation code named COCAGNE. This code relies on DIABOLO, a Simplified PN (SPN) method to compute the neutron flux inside the core for eigenvalue calculations. In order to assess the accuracy of SPN calculations, we have developed DOMINO, a new 3D Cartesian SN solver. The parallel implementation of DOMINO is very efficient and allows to complete an eigenvalue calculation involving around 300 x 10{sup 9} degrees of freedom within a few hours on a single shared-memory supercomputing node. This computation corresponds to a 26-group S{sub 8} 3D PWR core model used to assess the SPN accuracy. At the pin level, the maximal error for the SP{sub 5} DIABOLO fission production rate is lower than 0.2% compared to the S{sub 8} DOMINO reference for this 3D PWR core model. (authors)

  4. Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

    Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: luigi.debarberis@cec.eu.int; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)

    2006-04-15

    It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

  5. Anti -corrosion Effect of ETA on Materials in Secondary Loop of PWR

    2002-01-01

    In the world, over sixty percent of nuclear power plant have used advanced amunes ETA(Ethanolamine) as pH control agent in secondary loop of PWR. There are eighty percent of nuclear powerplants using ETA in USA. The corrosion of materials in steam generator (SG) tube and secondary looppower water reactor have been inhibited, the life of SG and the economics of the plant are increasedbecause of using ETA.

  6. Modeling and Simulation of Release of Radiation in Flow Blockage Accident for Two Loops PWR

    Khurram Mehboob; Cao Xinrong; Majid Ali

    2012-01-01

    In this study modeling and simulation of release of radiation form two loops PWR has been carried out for flow blockage accident. For this purpose, a MATLAB based program “Source Term Evaluator for Flow Blockage Accident” (STEFBA) has been developed, which uses the core inventory as its primary input. The TMI-2 reactor is considered as the reference plant for this study. For 1100 reactor operation days, the core inventory has been evaluated under the core design constrains at average reactor ...

  7. Chemical and radiochemical specifications - PWR power plants; Specifications chimiques et radiochimiques - Centrales REP

    Stutzmann, A. [Electricite de France (EDF), 93 - Saint-Denis (France)

    1997-07-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  8. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    Ade, Brian J [ORNL; Gauld, Ian C [ORNL

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

  9. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

    2014-06-01

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  10. PWR ENDF/B-VII cross-section libraries for ORIGEN-ARP

    McGraw, C. [Dept. of Nuclear Engineering, Texas A and M Univ., 3133 TAMU, College Station, TX 77843-3133 (United States); Ilas, G. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6172 (United States)

    2012-07-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% {sup 235}U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time. (authors)

  11. Evaluation of the RELAP4/MOD6 thermal-hydraulic code. [PWR

    Haigh, W.S.; Margolis, S.G.; Rice, R.E.

    1978-01-01

    The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA.

  12. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  13. NPP-VIIRS DNB-based reallocating subpopulations to mercury in Urumqi city cluster, central Asia

    Zhou, X.; Feng, X. B.; Dai, W.; Li, P.; Ju, C. Y.; Bao, Z. D.; Han, Y. L.

    2017-02-01

    Accurate and update assignment of population-related environmental matters onto fine grid cells in oasis cities of arid areas remains challenging. We present the approach based on Suomi National Polar-orbiting Partnership (S-NPP) -Visible Infrared Imaging Radiometer Suite (VIIRS) Day/Night Band (DNB) to reallocate population onto a regular finer surface. The number of potential population to the mercury were reallocated onto 0.1x0.1 km reference grid in Urumqi city cluster of China’s Xinjiang, central Asia. The result of Monte Carlo modelling indicated that the range of 0.5 to 2.4 million people was reliable. The study highlights that the NPP-VIIRS DNB-based multi-layered, dasymetric, spatial method enhances our abilities to remotely estimate the distribution and size of target population at the street-level scale and has the potential to transform control strategies for epidemiology, public policy and other socioeconomic fields.

  14. Estimation of damage by inmates of a PWR Reactor neutron irradiation. Project ZIRP; Estimacion del Dano por Irradiacion Neutronica en los Internos de un Reactor PWR. Proyecto ZIRP

    Cadenas Mendicoa, A. M.

    2013-07-01

    The study presented here focuses on the analysis of neutron and gamma irradiation damage suffered by the inmates of the JC NPP reactor metallic materials throughout its operational life. Such analysis of radiation are part of a project of great international impact, led by EPRI (Electric Power Research Institute) from the MRP (Materials Reliability Program), which aims to relate the degradation of the properties of metallic materials of the inmates of the reactor, with the conditions of operation and irradiation to which have been subjected during the operational life of the plant.

  15. Development of requirements tracking and verification technology for the NPP software

    Jung, Chul Hwan; Kim, Jang Yeol; Lee, Jang Soo; Song, Soon Ja; Lee, Dong Young; Kwon, Kee Choon

    1998-12-30

    Searched and analyzed the technology of requirements engineering in the areas of aerospace and defense industry, medical industry and nuclear industry. Summarized the status of tools for the software design and requirements management. Analyzed the software design methodology for the safety software of NPP. Development of the design requirements for the requirements tracking and verification system. Development of the background technology to design the prototype tool for the requirements tracking and verification.

  16. Deterministic and Probabilistic Analysis of NPP Communication Bridge Resistance Due to Extreme Loads

    Králik Juraj

    2014-12-01

    Full Text Available This paper presents the experiences from the deterministic and probability analysis of the reliability of communication bridge structure resistance due to extreme loads - wind and earthquake. On the example of the steel bridge between two NPP buildings is considered the efficiency of the bracing systems. The advantages and disadvantages of the deterministic and probabilistic analysis of the structure resistance are discussed. The advantages of the utilization the LHS method to analyze the safety and reliability of the structures is presented

  17. RADIATION HYGIENIC CONSEQUENCES OF THE ACCIDENT AT THE CHERNOBYL NPP AND THE TASKS OF THEIR MINIMIZATION

    G. G. Onischenko

    2009-01-01

    Full Text Available The paper presents data on the role and results of activities of Rospotrebnadzor bodies and institutions in the field of ensuring population radiation protection during various periods since accident at the Chernobyl NPP. Radiation hygienic characterization of territories affected by radioactive contamination from the accident, population exposure dose range, issues of ensuring radiological well-being of population and ways of their solution are being presented in the paper.

  18. Optimization of the energy complex “NPP-accumulator” in case of force majeure

    Zaluzhnaya, G.; Zagrebaev, A.

    2017-01-01

    We consider a problem of optimization of NPP with accumulator operation mode in case of force majeure. A mathematical formulation and solving of problem of energy output’s time behavior is provided. A mathematical formulation and solving of problem of energy’s optimum allocation to consumers with different priorities. Mathematically, the problem reduces to linear programming problem. We received that optimal time behavior is uniform energy output, and one should start with consumer with highest priority.

  19. Floor response spectra for seismic qualification of Kozloduy VVER 440-230 NPP

    Kostov, M.K. [Bulgarian Academy of Sciences, Sofia (BG). Central Lab. for Seismic Mechanics and Earthquake Engineering; Ma, D.C. [Argonne National Lab., IL (United States); Prato, C.A. [Univ. of Cordoba (AR); Stevenson, J.D. [Stevenson and Associates, Cleveland, OH (US)

    1993-08-01

    In this paper the floor response spectra generation methodology for Kozloduy NPP, Unit 1-2 of VVER 440-230 is presented. The 2D coupled soil-structure interaction models are used combined with a simplified correction of the final results for accounting of torsional effects. Both time history and direct approach for in-structure spectra generation are used and discussion of results is made.

  20. Computer modeling and simulators as part of university training for NPP operating personnel

    Volman, M.

    2017-01-01

    This paper considers aspects of a program for training future nuclear power plant personnel developed by the NPP Department of Ivanovo State Power Engineering University. Computer modeling is used for numerical experiments on the kinetics of nuclear reactors in Mathcad. Simulation modeling is carried out on the computer and full-scale simulator of water-cooled power reactor for the simulation of neutron-physical reactor measurements and the start-up - shutdown process.

  1. Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

    Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

    2005-03-15

    The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments.

  2. Integrated risk assessment for multi-unit NPP sites—A comparison

    Kumar, C. Senthil, E-mail: cskumar@igcar.gov.in [AERB-Safety Research Institute, Kalpakkam (India); Hassija, Varun; Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Balasubramaniyan, V. [AERB-Safety Research Institute, Kalpakkam (India)

    2015-11-15

    Highlights: • Framework for integrated risk assessment for multi-unit NPP sites. • Categorization of external and internal events. • Modelling of key issues: mission time, cliff-edge, common cause failures, etc. • Safety goals for multi-unit NPP sites. • Comparison of site core damage frequency in one, two, three and four unit sites. - Abstract: Most of the nuclear power producing sites in the world houses multiple units. Such sites are faced with hazards generated from external events: earthquake, tsunami, flood, etc. and can threaten the safety of nuclear power plants. Further, risk from a multiple unit site and its impact on the public and environment was evident during the Fukushima nuclear disaster in March 2011. It is therefore important to evolve a methodology to systematically assess the risk from multi-unit site. For a single unit site, probabilistic risk assessment technique identifies the potential accident scenarios, their consequences, and estimates the core damage frequency that arise due to internal and external hazards. This challenging task becomes even more complex for a multiple unit site, especially when the external hazards that has the potential to generate one or more correlated hazards or a combination of non-correlated hazards are to be modelled. This paper presents an approach to evaluate risk for multiple NPP sites and also compare the risk for sites housing single, double and multiple nuclear plants.

  3. The Fukushima Dai-ichi NPP accident crisis and its influence on energy policy in Japan

    Uchida, Shunsuke; Naitoh, Masanori; Okada, Hidetoshi; Suzuki, Hiroaki [Institute of Applied Energy, Tokyo (Japan)

    2011-10-15

    Fossil fuel power plants (FPPs) and nuclear power plants (NPPs) along the northern Pacific coast of Japan experienced a mega earthquake and resulting tsunamis on March 11, 2011, which resulted in serious damage. More than half of the FPPs have returned to supplying electricity, while all NPPs are still shut down. In particular, Fukushima Dai-ichi Nuclear Power Plant (NPP) of the Tokyo Electric Power Co. experienced a nuclear accident crisis unprecedented in both scale and timeframe. This paper gives a brief overview of the events and their propagation based on the International Atomic Energy Agency (IAEA) report and the authors' studies on accident analysis, and offers considerations on root causes of the propagation, e.g., problems in hardware, software and accident management, by comparing the responses of Fukushima Dai-ichi NPP with those of other reactors at the Fukushima Dai-ni, Tokai, Onagawa and Higashidori NPPs, where cold shutdowns were successfully maintained even though they had also been affected by the earthquake and tsunamis. Future technical subjects for safe NPP operation and the influence of the events on Japanese energy policy are presented. (orig.)

  4. Conformity Between LR0 Mock-Ups and Vvers Npp Rpv Neutron Flux Attenuation

    Belousov, Sergey; Ilieva, Krassimira; Kirilova, Desislava

    2009-08-01

    The conformity of the mock-up results and those for reactor pressure vessel (RPV) of nuclear power plants (NPP) has been evaluated in order to qualify if the mock-ups data could be used for benchmark's purpose only, or/and for simulating of the NPP irradiation conditions. Neutron transport through the vessel has been calculated by the three-dimensional discrete ordinate code TORT with problem oriented multigroup energy neutron cross-section library BGL. Neutron flux/fluence and spectrum shape represented by normalized group neutron fluxes in the multigroup energy structure, for neutrons with energy above 0.5 MeV, have been used for conformity analysis. It has been demonstrated that the relative difference of the attenuation factor as well as the group neutron fluxes did not exceed 10% at all considered positions for VVER-440. For VVER-1000, it has been obtained the same consistency, except for the location behind the RPV. The neutron flux attenuation behind the RPV is 18% higher than the mock-up attenuation. It has been shown that this difference arises from the dissimilarity of the biological shielding. The obtained results have demonstrated that the VVERs' mock-ups are appropriate for simulating the NPP irradiation conditions. The mock-up results for VVER-1000 have to be applied more carefully i.e. taking into account the existing peculiarity of the biological shielding and RPV attenuation azimuthal dependence.

  5. Development of seismic safety reevaluation procedure considering the ageing of NPP facilities

    Lee, Myoung Kue [Jeonju Univ., Cheonju (Korea, Republic of); Kim, J. M. [Cheonnam National Univ., Gwangju (Korea, Republic of); Kim, Y. S.; Cheong, S. H.; Kim, I. S.; Lee, M. G.; Kim, D. O. [Andong National Univ., Andong (Korea, Republic of); Lee, G. H. [Mokpo National Maritime Univ., Mokpo (Korea, Republic of)

    2003-03-15

    There are three of Nuclear Power Plants subject to the USI A-46 in Korea, including Kori No 1 and No 2 and Wolsung No 1. For the sake of resolution of the issue the possibility of adopting the GIP developed by the SQUG in USA is very high. In relation to the issue, this study addresses some technical improvements of the GIP including sloshing analysis based on multiple modes, seismic retrofit of cabinet for reduction of ICRS and modification of IRS depending on damping ratio. Dominant degradation factor and its affects NPP concrete elements are reviewed : chloride induced corrosion, carbonation of concrete elements, freezing and thawing of concrete elements, chemical and biological process, crack affect on concrete degradation. Various technical reports and papers about age-related degradation are reviewed for identification of degradation properties of NPP structures and components and degradation trend in NPP structures and components. This report summarizes numerical model for concrete degradation and development procedure of numerical models for concrete degradation. This report proposes the research necessity for performance evaluation of degraded concrete structure and selection of element for further study.

  6. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  7. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Jimenez, G.; Martinez, R. M.; Fernandez, K.; Morato, D. J.; Bocanegra Melian, R.; Mena, L.; Queral, C.

    2014-07-01

    During the development of a severe accident in a PWR reactor can be generated, large quantities of hydrogen by the oxidation of metals present in the nucleus, mainly the zirconium fuel pods. This hydrogen, along with steam and other gases, can be released to the atmosphere of contention by a leak or break in the primary circuit and achieving conditions in which combustion may occur. Combustion causes thermal and pressure loads that can damage the security systems and the integrity of the containment building, last barrier of confinement of radioactive materials. The main condition that defines the characteristics of the combustion is the concentration of species, detailed knowledge of the distribution of hydrogen is very important to correctly predict the possible damage to the containment in the event that there is combustion. (Author)

  8. Optimization of thermal efficiency of nuclear central power like as PWR; Otimizacao da eficiencia termica de uma usina nuclear do tipo PWR

    Lapa, Nelbia da Silva

    2005-10-15

    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

  9. Valve inlet fluid conditions for pressurizer safety and relief valves in Westinghouse-designed plants. Final report. [PWR

    Meliksetian, A.; Sklencar, A.M.

    1982-12-01

    The overpressure transients for Westinghouse-designed NSSSs are reviewed to determine the fluid conditions at the inlet to the PORV and safety valves. The transients considered are: licensing (FSAR) transients; extended operation of high pressure safety injection system; and cold overpressurization. The results of this review, presented in the form of tables and graphs, define the range of fluid conditions expected at the inlet to pressurized safety and power-operated relief valves utilized in Westinghouse-designed PWR units. These results will provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI/PWR Safety and Relief Valve Test Program indeed envelop those expected in their units.

  10. 浅议福岛核电事故对我国核电发展的影响及借鉴%Lessons from Fukushima Dai-ichi NPP Accident and Effect on NPP Development in China

    潘金钊

    2012-01-01

    Although the Fukushima NPP accident caused no serious environmental problem in China, but it does have great effect on the future NPP development in China. The NPP development plan in China is so uniquely huge in the world energy field that great challenge has already arisen to the Chinese NPP constructors and operators. This paper analyzes the sequence of the Fukushima NPP, which can be referred to in the improvement of the NPP development plan and the NPP accident emergency plan.%日本福岛核电事故虽然未对我国的环境造成严重影响,但是该事故的发生对我国后续核电的发展必然产生重大影响.我国庞大、密集的核电发展规划在世界能源发展领域是绝无仅有的,无论是在技术路线、标准制订还是在建造、运行的组织管理上,我国核电建设者和管理者都将面临巨大考验.通过分析和借鉴本次日本福岛核电事故的发生、处理过程,将对我国核电发展在多方面提供重要参考,从而促进我国核电规划及核电事故应急体系的持续完善.

  11. Effect of co-free valve on activity reduction in PWR

    Bahn, C.B.; Han, B.C.; Bum, J.S.; Hwang, I.S. [Department of Nuclear Engineering, Seoul National Univ. (Korea, Republic of); Lee, C.B. [Korea Atomic Energy Research Inst., Daejon (Korea, Republic of)

    2002-07-01

    Radioactive nuclei, such as {sup 68}Co and {sup 60}Co, deposited on out-of-core surfaces in a pressurized water reactor (PWR) primary coolant system, are major sources of occupational radiation exposure to plant maintenance personnel and act as costly impediment to prompt and effective repairs. Valve hardfacing alloys exposed to primary coolant are considered as one of the main Co sources. To evaluate the Co-free valve, such as NOREM 02 and Deloro 50, the candidates for the alternative to Stellite 6, in a simulated PWR primary condition, SNU corrosion test loop (SCOTL) was constructed. For gate valves hard-faced with made of NOREM 02 and Deloro 50 hot cycling tests were conducted for up to 2,000 on-off cycles with cold leak tests at 1,000 cycle interval. It was observed that the leak rate of NOREM 02 (Fe-base) did not satisfy the nuclear grade valve leak criteria. After 1000 cycles test, while there was no leakage in case of Deloro 50 (Ni-base). Also, Deloro 50 showed no leakage after 2000 cycles. To estimate the activity reduction effect, we modified CRUDSIM-MIT which modeled the effects of coolant chemistry on the crud transport and activity buildup in the primary system of PWR. In the new code, crud evaluation and assessment (CREAT), {sup 60}Co activity buildup prediction includes 1) Co-base valve replacement effect, 2) Co-base valve maintenance effect, and 3) control rod drive mechanism (CRDM) and main coolant pump (MCP) shaft contribution. CREAT predicted that the main contributor of Co activity buildup was the corrosion-induced release of Co from the steam generator (SG) tubing. With new SG's tubed with alloy 690, Korean Next Generation Reactor (APR-1400) is expected to have about 64% lower Co activity on SG surface. The use of all Co-free valves is expected to cut additional 8% of activity which is only marginal. (authors)

  12. PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF

    Benfarah Moez

    2016-01-01

    Full Text Available Normal operation of PWR generates corrosion and wear products in the primary circuit which are activated in the core and constitute the major source of the radiation field. In addition, cases of fuel failure and alpha emitter dissemination in the coolant system could represent a significant radiological risk. Radiation field and alpha risks are the main constraints to carry out maintenance and to handle effluents. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products and actinides and to carry out the appropriate measurements in PWR circuits and loop experiments. As a matter of fact, it is more than necessary to develop and use a reactor contamination assessment code in order to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has actually been developed and used for this aim. It is presented in this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination. Then, the use of OSCAR is illustrated with two examples from our engineering studies. The first example of OSCAR engineering studies is linked to the behavior of the activated corrosion products. The selected example carefully explores the impact of the restart conditions following a reactor mid-cycle shutdown on circuit contamination. The second example of OSCAR use concerns fission products and disseminated fissile material behavior in the primary coolant. This example is a parametric study of the correlation between the quantity of disseminated fuel and the variation of Iodine 134 in the primary coolant.

  13. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  14. Evolution of reactor monitoring and protection systems for PWR; Evolution des systemes de surveillance et de protection des REP

    Chaloin, B. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Mourlevat, J.L. [FRAMATOME ANP, 92 - Paris-La-Defence (France)

    2004-07-01

    This paper presents the evolution of the reactor protection systems and of the reactor monitoring systems for PWR since the initial design in the Fessenheim plant to the latest development for the EPR (European pressurized reactor). The features of both systems for the different kinds of PWR operating in France: 900 MWe, 1300 MWe and N4, are reviewed. The expected development of powerful micro-processors for computation, for data analysis and data storage will make possible in a near future the monitoring on a 3-dimensional basis and on a continuous manner, of the nuclear power released in the core. (A.C.)

  15. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

    2016-10-15

    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

  16. Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT

    Uwe Imke

    2012-01-01

    Full Text Available SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT. The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy.

  17. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code. [PWR

    Ramsthaler, J. A.; Lime, J. F.; Sahota, M. S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A.

  18. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  19. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  20. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

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

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  1. Use of plutonium in PWR-type reactors; Utilisation du plutonium dans les REP

    Berthet, A. [Electricite de France (EDF), 75 - Paris (France). Direction de l' Equipement

    1999-04-01

    The plutonium is used, as fuel, in the pressurized water reactors. It does not exist in nature; butit is fabricated in the reactor by neutrons capture. The MOX (Mixed Oxides) is its usual name. A part is consumed by the fission, the remainder is found in the used fuel released from the reactor. The paper deals with the plutonium specificities, the research and development programs about this fuel. The technical specifications of the PWR recycling the plutonium are also included (radiation protection, reactor fueling). (A.L.B.)

  2. Application of LBB to high energy piping systems in operating PWR

    Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  3. Assessment Impacts of Weather and Land Use/Land Cover (LULC Change on Urban Vegetation Net Primary Productivity (NPP: A Case Study in Guangzhou, China

    Xiantie Zeng

    2013-08-01

    Full Text Available Net primary productivity (NPP can indicate vegetation ecosystem services ability and reflect variation response to climate change and human activities. This study applied MODIS-1 km NPP products to investigate the NPP variation from 2001 to 2006, a fast urban expansion and adjustment period in Guangzhou, China, and quantify the impacts of weather and land use/land cover (LULC changes, respectively. The results showed that the NPP mean value increased at a rate of 11.6 g∙C∙m−2∙yr−1 during the initial three years and decreased at an accelerated rate of 31.0 g∙C∙m−2∙yr−1 during the final three years, resulting in a total NPP loss of approximately 167 × 106 g∙C. The spatiotemporal of NPP varied obviously in the central area, suburb and exurb of Guangzhou driven by three patterns of weather and LULC changes. By the interactive effects and the weather variation dominated effects, NPP of most areas changed slightly with dynamic index less than 5% of NPP mean value in the central area and the suburb. The LULC change dominated effects caused obvious NPP reduction, by more than 15% of the NPP mean value, which occurred in some areas of the suburb and extended to the exurb with the outward urban sprawl. Importantly, conversion from wood grassland, shrublands and even forests to croplands occupied by urban landscapes proved to be a main process in the conversion from high-NPP coverage to low-NPP coverage, thereby leading to the rapid degradation of urban carbon stock capacity in urban fringe areas. It is helpful for government to monitor urban ecological health and safety and make relevant policies.

  4. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  5. Estimate of the speed of the refrigerant on a PWR: three way based on the analysis of noise; Estimacion de la volecidad del refrigerante en un PWR: tres vias basadas en el analisis de ruido

    Montalvo, C.; Ruiz, M.; Garcia Berrocal, A.

    2014-07-01

    The speed of the refrigerant is a key parameter in the monitoring of the operation a PWR. He know this value and be able to track on-site It allows an understanding of the State of the kernel with valuable information about the refrigerant, and thus behavior on heat exchange which takes place in the reactor. (Author)

  6. CFD analysis of the temperature field in emergency pump room in Loviisa NPP

    Rämä, Tommi, E-mail: tommi.rama@fortum.com [Fortum Power and Heat, P.O.B. 100, FI-00048 Fortum (Finland); Toppila, Timo, E-mail: timo.toppila@fortum.com [Fortum Power and Heat, P.O.B. 100, FI-00048 Fortum (Finland); Kelavirta, Teemu, E-mail: teemu.kelavirta@fortum.com [Fortum Power and Heat, Loviisa Power Plant, P.O.B. 23, FI-07901 Loviisa (Finland); Martin, Pasi, E-mail: pasi.martin@fortum.com [Fortum Power and Heat, Loviisa Power Plant, P.O.B. 23, FI-07901 Loviisa (Finland)

    2014-11-15

    Highlights: • Laser scanned room geometry from Loviisa NPP was utilized for CFD simulation. • Uncertainty of CFD simulation was estimated using the Grid Convergence Index. • Measured temperature field of pump room was reproduced with CFD simulation. - Abstract: In the Loviisa Nuclear Power Plant (NPP) six emergency pumps belonging to the same redundancy are located in the same room. During a postulated accident the cooling of the room is needed as the engines of the emergency pumps generate heat. Cooling is performed with fans blowing air to the upper part of the room. Temperature limits have been given to the operating conditions of the main components in order to ensure their reliable operation. Therefore the temperature field of the room is important to know. Temperature measurements were made close to the most important components of the pump room to get a better understanding of the temperature field. For these measurements emergency pumps and cooling fan units were activated. To simulate conditions during a postulated accident additional warm-air heaters were used. Computational fluid dynamic (CFD) simulations were made to support plant measurements. For the CFD study one of the pump rooms of Loviisa NPP was scanned with a laser and this data converted to detailed 3-D geometry. Tetrahedral computation grid was created inside the geometry. Grid sensitivity studies were made, and the model was then validated against the power plant tests. With CFD the detailed temperature and flow fields of the whole room were produced. The used CFD model was able to reproduce the temperature field of the measurements. Two postulated accident cases were simulated. In the cases the operating cooling units were varied. The temperature profile of the room changes significantly depending on which units are cooling and which only circulating the air. The room average temperature stays approximately the same. The simulation results were used to ensure the acceptable operating

  7. OVERVIEW ON BNL ASSESSMENT OF SEISMIC ANALYSIS METHODS FOR DEEPLY EMBEDDED NPP STRUCTURES.

    XU,J.; COSTANTINO, C.; HOFMAYER, C.; GRAVES, H.

    2007-04-01

    A study was performed by Brookhaven National Laboratory (BNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (USNRC), to determine the applicability of established soil-structure interaction analysis methods and computer programs to deeply embedded and/or buried (DEB) nuclear power plant (NPP) structures. This paper provides an overview of the BNL study including a description and discussions of analyses performed to assess relative performance of various SSI analysis methods typically applied to NPP structures, as well as the importance of interface modeling for DEB structures. There are four main elements contained in the BNL study: (1) Review and evaluation of existing seismic design practice, (2) Assessment of simplified vs. detailed methods for SSI in-structure response spectrum analysis of DEB structures, (3) Assessment of methods for computing seismic induced earth pressures on DEB structures, and (4) Development of the criteria for benchmark problems which could be used for validating computer programs for computing seismic responses of DEB NPP structures. The BNL study concluded that the equivalent linear SSI methods, including both simplified and detailed approaches, can be extended to DEB structures and produce acceptable SSI response calculations, provided that the SSI response induced by the ground motion is very much within the linear regime or the non-linear effect is not anticipated to control the SSI response parameters. The BNL study also revealed that the response calculation is sensitive to the modeling assumptions made for the soil/structure interface and application of a particular material model for the soil.

  8. MetEd Resources for Embracing Advances with S-NPP and JPSS

    Abshire, W. E.; Dills, P. N.; Weingroff, M.

    2014-12-01

    The COMET® Program (www.comet.ucar.edu), a part of the UCAR Community Programs (UCP) at UCAR, receives funding from NOAA NESDIS as well as EUMETSAT and the Meteorological Service of Canada to support education and training in satellite meteorology. For many years COMET's satellite education programs have focused on developing self-paced online educational materials that highlight the capabilities and applications of current and next-generation operational geostationary and polar-orbiting satellites and their relevance to operational forecasters and other user communities. By partnering with experts from the Naval Research Laboratory, NOAA-NESDIS and its Cooperative Institutes, Meteorological Service of Canada, EUMETSAT, and other user communities, COMET stimulates greater use of current and future satellite observations and products. This presentation provides a tour of COMET's satellite training and education offerings that are directly applicable to data and products from the S-NPP and JPSS satellite series. A recommended set of lessons for users who wish to learn more will be highlighted, including excerpts from the newest materials on the Suomi NPP VIIRS imager and its applications, as well as advances in nighttime visible observation with the VIIRS Day-Night Band. We'll show how the lessons introduce users to the advances these systems bring to forecasting, numerical weather prediction, and environmental monitoring. Over 90 satellite-focused, self-paced, online materials are freely available on the of the MetEd Web site (http://www.meted.ucar.edu) via the "Education & Training", "Satellite" topic area. Quite a few polar-orbiting-related lessons are available in both English, Spanish, and French. Additionally, S-NPP and JPSS relevant information can also be found on the the Environmental Satellite Resource Center (ESRC) Web site (www.meted.ucar.edu/esrc) that is maintained by COMET. The ESRC is a searchable, database-driven Web site that provides access to

  9. Lunar BRDF Correction of Suomi-NPP VIIRS Day/Night Band Time Series Product

    Wang, Z.; Roman, M. O.; Kalb, V.; Stokes, E.; Miller, S. D.

    2015-12-01

    Since the first-light images from the Suomi-NPP VIIRS low-light visible Day/Night Band (DNB) sensor were received in November 2011, the NASA Suomi-NPP Land Science Investigator Processing System (SIPS) has focused on evaluating this new capability for quantitative science applications, as well as developing and testing refined algorithms to meet operational and Land science research needs. While many promising DNB applications have been developed since the Suomi-NPP launch, most studies to-date have been limited by the traditional qualitative image display and spatial-temporal aggregated statistical analysis methods inherent in current heritage algorithms. This has resulted in strong interest for a new generation of science-quality products that can be used to monitor both the magnitude and signature of nighttime phenomena and anthropogenic sources of light emissions. In one particular case study, Román and Stokes (2015) demonstrated that tracking daily dynamic DNB radiances can provide valuable information about the character of the human activities and behaviors that influence energy, consumption, and vulnerability. Here we develop and evaluate a new suite of DNB science-quality algorithms that can exclude a primary source of background noise: i.e., the Lunar BRDF (Bidirectional Reflectance Distribution Function) effect. Every day, the operational NASA Land SIPS DNB algorithm makes use of 16 days worth of DNB-derived surface reflectances (SR) (based on the heritage MODIS SR algorithm) and a semiempirical kernel-driven bidirectional reflectance model to determine a global set of parameters describing the BRDF of the land surface. The nighttime period of interest is heavily weighted as a function of observation coverage. These gridded parameters, combined with Miller and Turner's [2009] top-of-atmosphere spectral irradiance model, are then used to determine the DNB's lunar radiance contribution at any point in time and under specific illumination conditions.

  10. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    Freire, Luciano Ondir, E-mail: luciano.ondir@gmail.com; Andrade, Delvonei Alves de, E-mail: delvonei@ipen.br

    2014-12-15

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs.

  11. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  12. Construction and utilization of linear empirical core models for PWR in-core fuel management

    Okafor, K.C.

    1988-01-01

    An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k{sub {infinity}} profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results.

  13. An extension of the validation of SCALE (SAS2H) isotopic predictions for PWR spent fuel

    DeHart, M.D.; Hermann, O.W.

    1996-09-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms. Unlike fresh fuel assumptions typically used in criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict spent fuel composition; these isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the depletion codes and data, experiment is compared with predictions; such comparisons have been done in earlier ORNL work. This report describes additional independent measurements and corresponding calculations as a supplement. The current work includes measured isotopic data from 19 spent fuel samples from the Italian Trino Vercelles PWR and the US Turkey Point-3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results on combination of measured-to-calculated ratios are presented. The results described herein represent an extension to a new reactor design and spent fuel samples with enrichment as high as 3.9 wt% {sup 235}U. Consistency with the earlier work for each of two different cross-section libraries suggests that the estimated biases for each of the isotopes in the earlier work are reasonably good estimates.

  14. PWR-FBR with closed fuel cycle for a sustainable nuclear energy supply in China

    XU Mi

    2007-01-01

    From the thermal reactor to the fast reactor and then to the fusion reactor; this is the three-step strategy that has been decided for a sustainable nuclear energy supply in China. As the main thermal reactor type, the commercialized development phase of the pressurized water reactor (PWR) has been stepped up. The development of the fast reactor (FBR) is still in the early stage, marked by China experimental fast reactor (CEFR), which is currently under construction. According to the strategy study on the fast reactor development in China, its engineering development will be divided into three steps: the CEFR with a power of 65 MWt 20 Mwe; the China prototype fast reactor (CPFR) with a power of 1 500 MWt/600 Mwe; and the China demonstration fast reactor (CDFR) with a power of 2 500-3 750 MWt 1 000-1 500 Mwe. With regards to the fuel cycle, a 100 ta PWR spent fuel reprocessing pilot plant and a 500 kg/a MOX fabrication plant are under construction. A project involving the construction of an industrial reprocessing plant and an MOX fabrication plant are also under application phase.

  15. Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test). [PWR

    None

    1977-06-01

    Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations.

  16. Accelerated IGA/SCC testing of Alloy 600 in contaminated PWR environments

    Miglin, B.P.; Sarver, J.M. [Babcock & Wilcox R& D Division, Alliance, OH (United States); Aoki, K. [NFI, Osaka (Japan); Koch, D.W. [Babcock & Wilcox Nuclear Services, Lynchburg, VA (United States); Takamatsu, H. [Kansai Electric, Osaka (Japan)

    1992-12-31

    An accelerated corrosion test (360{degrees}C for 2000 hrs) was performed on C-ring specimens machined from one heat of Alloy 600 tubing in the mill-annealed condition. The specimens were exposed to secondary-side pressurized-water-reactor (PWR) solutions contaminated with lead, sulfur, silicon, and a combination of these contaminants. Where possible, MULTEQ calculations were performed to determine the chemical concentrations so that a constant elevated-temperature pH of 4.5 was achieved. This test was designed to examine the ability of these contaminants to cause intergranular attack and/or stress corrosion in stressed Alloy 600 tubing. The results from this test demonstrated that under the test conditions used, lead-contaminated PWR secondary water induces and propagates intergranular attack (IGA) and stress corrosion cracking (SCC) in Alloy 600. Attack was intergranular; the degree of attack did not vary in the liquid or vapor portions of the test environments. Although attack was more severe at higher stresses, significant attack was observed in samples stressed to the typical operating stress. Solutions of only sulfur and only silicon displayed no initiation or propagation of either IGA or SCC. However, the solution containing all three contaminants caused attack with identical morphology to that observed in the lead-contaminated solution.

  17. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D. [Korea Atomic Energy Research Inst., Daejeon (Korea); Choi, B.I.; Lee, H.Y.; Song, M.J. [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea)

    2004-07-01

    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 {open_square} under the normal condition. The off-normal condition has an environmental temperature of 40 {open_square}. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions.

  18. Vulnerability of a partially flooded PWR reactor cavity to a steam explosion

    Cizelj, Leon [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia)]. E-mail: leon.cizelj@ijs.si; Koncar, Bostjan [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia); Leskovar, Matjaz [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia)

    2006-08-15

    When the hot molten core comes into contact with the water in the reactor cavity a steam explosion may occur. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and later, during the expansion of the water vapour, to production of missiles that may endanger surrounding structures. The purpose of the performed analysis is to provide an estimation of the expected pressure loadings on the typical PWR cavity structures during a steam explosion, and to make an assessment of the vulnerabilities of the typical PWR cavity structures to steam explosions. To achieve this, the fit-for-purpose steam explosion model is proposed, followed by comprehensive and reasonably conservative analyses of two typical ex-vessel steam explosion cases differing in the steam explosion energy conversion ratio. In particular, the vulnerability of the surrounding reinforced concrete walls to damage has been sought in both cases.

  19. PWR Containment Shielding Calculations with SCALE6.1 Using Hybrid Deterministic-Stochastic Methodology

    Mario Matijević

    2016-01-01

    Full Text Available The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS were demonstrated when applied to a realistic deep penetration Monte Carlo (MC shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR parameters is based on deterministic transport theory (SN method providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence between SN-MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources.

  20. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)

    2015-08-15

    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

  1. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, Saclay (France); Mimouni, S., E-mail: stephane.mimouni@edf.fr [Electricité de France, EDF MF2E, Chatou (France); Manzini, G., E-mail: giovanni.manzini@rse-web.it [RSE, Milano (Italy); Xiao, J., E-mail: jianjun.xiao@kit.edu [IKET, KIT, Karlsruhe (Germany); Vyskocil, L., E-mail: vyl@ujv.cz [UJV Rez (Czech Republic); Siccama, N.B., E-mail: siccama@nrg.eu [NRG, Safety and Power (Netherlands); Huhtanen, R., E-mail: risto.huhtanen@vtt.fi [VTT, PO Box 1000, FI-02044 VTT (Finland)

    2015-02-15

    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.

  2. Surface Oxidation Phenomena of Ni-Based Alloy 600 in PWR Primary Water Conditions

    Lim, Yun Soo; Hwang, Seong Sik; Kim, Sung Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    There is, nevertheless, growing evidence in support for the internal oxidation model by Scot, in which grain boundary oxidation is responsible for embrittlement and cracking. Grain boundaries can act as an enhanced diffusion path for oxidation, and grain boundary oxidation can be regarded as a precursor for crack initiation. Oxidation of the grain boundary in almost all nickel-based alloys exposed to primary water is known to be detrimental for grin boundary cohesion. Panter et al. showed that the crack initiation time is strongly reduced when the specimens are pre-exposed in a simulated PWR environment in the absence of applied stress. The changes of the grain boundary structure and chemistry owing to oxygen penetration can increase the sensitivity to PWSCC under a load since grain boundary oxidization significantly weakens the grain boundary strength. Most of the important experimental results obtained are believed to correlate with the oxidation penetration into the material. A spinel structure was detected by XRD in the oxide layers. Several different types of oxide scales were found by SEM examination on the corroded surface of Alloy 600 after an immersion test in the primary water environments. Surface grain boundaries were oxidized by oxygen penetration into the matrix through grain boundaries. Grain boundary oxidization is thought to be the main reason for intergranular cracking in this alloy in a primary water environment of a PWR.

  3. Consideration of MAAP 5.0.2 ESF Model Characteristics for APR1400 NPP

    Seo, Mi Ro; Kim, Hyeong Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    After the Fukushima accident, EPRI has developed the MAAP (Modular Accident Analysis Program) version 5 that is expected to make up the limitation of MAAP4, that is to say, the appropriateness of the model for the severe accident phenomena and the applicability to the phenomena in the spent fuel pool and the halfloop operation. Up to now, the newest version of MAAP is 5.0.2 (Build 5020000) that was released officially in December, 2013. In addition to this, it is expected that MAAP 5.0.3 version will be published sooner or later. As a kind of post-Fukushima measures, KHNP is developing the probabilistic safety assessment (PSA) and severe accident management guideline (SAMG) for low power and shutdown (LPSD) mode and MAAP 5.0.2 should be used in these projects as a major analysis program. So, first of all, it is necessary that the parameter file for domestic NPP should be upgraded as current Ver. MAAP4 to Ver. MAAP 5.0.2. KHNP has developed the draft version of parameter file for APR1400 type NPP and is being upgraded continuously. The Engineering Safety Features (ESF) model is one of the unique features of MAAP. In this study, we try to share the general information of the MAAP ESF model and the specific characteristics of APR1400 ESF model based on the newly developed MAAP 5.0.2 parameter file. Currently, while developing the LPSD PSA and LPSD SAMG as a kind of post-Fukushima measures, KHNP have the plan in order to upgrade the old parameter file based on MAAP4 to that based on MAAP5.0.2 for all domestic nuclear power plants. And, as the first effort, we are developing the MAAP 5.0.2 parameter file for APR1400 type NPP. In this study, we tried to develop the more accurate and reasonable ESF model of APR1400. In this process, we can find the distinctions and characteristics of specific ESF model and generalized ESF model of MAAP5.0.2. Also, we can eliminate the confusing concepts existed in the two models. So, it is judged that the newly developed MAAP5

  4. MORE: Management of requirements in NPP modernisation projects. Project report 2007

    Fredriksen, R.; Katta, V.; Raspotnig, C. [Institutt for energiteknikk (IFE) (Norway); Valkonen, J. [Technical Research Centre of Finland (VTT) (Finland)

    2008-03-15

    This report documents the work and related activities of the MORE project in the period January 1 - December 31 in 2007. The focus of this report is on improvements of the former project results, to identify and apply a couple of case studies from NPP projects, and activities in order to initiate and implement the industrial take-up and utilisation of the research results in real modernisation projects. The report also provides a brief description of the extended industrial network and disseminations of the results in Nordic and NKS related events such as seminars and workshops. (au)

  5. Inter-Comparison of NPP/CrIS with AIRS and IASI

    Wang, L.; Han, Y.; Weng, F.; Goldberg, M.

    2012-12-01

    The Cross-track Infrared Sounder (CrIS) on the newly-launched Suomi National Polar-orbiting Partnership (Suomi NPP) and future Joint Polar Satellite System (JPSS) is a Fourier transform spectrometer that provides soundings of the atmosphere with 1305 spectral channels, over 3 wavelength ranges: LWIR (9.14 - 15.38 μm); MWIR (5.71 - 8.26 μm); and SWIR (3.92 - 4.64 μm). An accurate spectral and radiometric calibration as well as geolocation is fundamental for CrIS radiance Sensor Data Records (SDRs). In this study, through inter- and intra-satellite calibration efforts, we focus on assessment of NPP/CrIS post-launch radiometric and spectral calibration. The purpose of this study is to use inter-calibration technologies to quantify the CrIS calibration bias and uncertainties. We will compare CrIS hyperspectral radiance measurements with the Atmospheric Infrared Sounder (AIRS) on NASA Earth Observing System (EOS) Aqua and Infrared Atmospheric Sounding Interferometer (IASI) on Metop-A and -B to examine spectral and radiometric consistence and difference among three hyperspectral IR sounders. The newly-launched CrIS on Suomi NPP, combined with AIRS and IASI, provide the first-ever inter-calibration opportunity because three hyperspectral IR sounders can observe the Earth and Atmosphere at the same spectral regions from different satellites. We will directly compare CrIS with AIRS and IASI at orbital crossing points of satellites occurring at high latitudes, the so-called simultaneous nadir overpasses (SNO). The CrIS, AIRS, and IASI spectra will be processed at common grids and then the spectral differences will be computed. In addition, an accurate collocation algorithm has been developed to collocate high spatial resolution measurements from the Visible Infrared Imager Radiometer Suite (VIIRS) within each CrIS Field of View (FOV). The collocated VIIRS radiances will be used to characterize the homogeneity of CrIS FOVs to further reduce comparison uncertainties

  6. Impact of thicker cladding on the nuclear parameters of the NPP Krsko fuel

    Kromar, Marjan, E-mail: marjan.kromar@ijs.s [Jozef Stefan Institute, Reactor Physics Department, Jamova 39, 1001 Ljubljana (Slovenia); Kurincic, Bojan [Nuclear Power Plant Krsko, Engineering Division, Nuclear Fuel and Reactor Core, Vrbina 12, 8270 Krsko (Slovenia)

    2011-04-15

    To make fuel rods more resistant to grid-to-rod fretting or other cladding penetration failures, the cladding thickness could be increased or strengthened. Implementation of thicker fuel rod cladding was evaluated for the NPP Krsko that uses 16 x 16 fuel design. Cladding thickness of the Westinghouse standard fuel design (STD) and optimized fuel design (OFA) is increased. The reactivity effect during the fuel burnup is determined. To obtain a complete realistic view of the fuel behaviour a typical, near equilibrium, 18-month fuel cycle is investigated. The most important nuclear core parameters such as critical boron concentrations, isothermal temperature coefficient and rod worth are determined and compared.

  7. Research on General Corrosion Property of 304L and 304NG Stainless Steels in Simulated PWR Primary Water

    PENG; De-quan; HU; Shi-lin; ZHANG; Ping-zhu; WANG; Hui

    2012-01-01

    <正>The general corrosion behaviors of 304L and 304NG grade stainless steels in simulated pressurized water reactor (PWR) primary loop were studied using still autoclave, respectively, the corrosion test lasted for 1 680 hours. The corrosion oxide films were analyzed macroscopically and microscopically. The results are shown in Figs. 1, 2.

  8. Evaluation of Passive Autocatalytic Recombiner (PAR) Implementation in a Konvoi NPP Containment Type

    Lopez-Alonso, E.; Papini, D.; Jimenez, G.

    2015-07-01

    The evaluation of Passive Autocatalytic Recombiner (PAR) implementation has been developed under the methodology extracted from the IAEA document, analysing the size, location and number of the PARs capable to minimize the combustion risk, which arises from a hydrogen release generated during a severe accident and its distribution in containment building. A detailed three-dimensional model of Konvoi (PWR) containment with GOTHIC 8.1 code was used for the simulations. The hydrogen preferential pathways and the accumulation points were studied and identified on the basis of a base case scenario without any mitigation measure. The PAR configuration offers an improvement in the chosen accidental scenario; decreasing the possibility of hydrogen combustion and leading to concentration values below the flammability limit (hydrogen concentration below 7%), in all the containment compartments at the end of the transient. Furthermore, from the analysis, it is concluded that the time required to reach hydrogen concentrations below the combustion limit is considerably reduced. (Author)

  9. Modelling the current and future spatial distribution of NPP in a Mediterranean watershed

    Donmez, Cenk; Berberoglu, Suha; Curran, Paul J.

    2011-06-01

    The aim of this study is to use full spatial resolution Envisat MERIS data to drive an ecosystem productivity model for pine forests along the Mediterranean coast of Turkey. The Carnegie, Ames, Stanford Approach (CASA) terrestrial biogeochemical model, designed to simulate the terrestrial carbon cycle using satellite sensor and meteorological data, was used to estimate annual regional fluxes in terrestrial net primary productivity (NPP). At its core this model is based on light-use efficiency, influenced by temperature, rainfall and solar radiation. Present climate data was generated from 50 climate stations within the watershed using co-kriging. Regional scale pseudo-warming data for year 2070 were derived using a Regional Climate Model (RCM) these data were used to downscale the GCM General Circulation Model for the research area as part of an international research project called Impact of Climate Changes on Agricultural Production Systems in Arid Areas (ICCAP). Outputs of climate data can be moderated using the four variables of percent tree cover, land cover, soil texture and NDVI. This study employed 47 MERIS images recorded between March 2003 and September 2005 to derive percent tree cover, land cover and NDVI. Envisat MERIS data hold great potential for estimating NPP with the CASA model because of the appropriateness of both its spatial and its spectral resolution.

  10. Radionuclide Distribution in the Soil on the Stabatishkes Site in the Vicinity of the Ignalina NPP

    Jevgenij Aliončik

    2011-02-01

    Full Text Available A near surface repository for low and intermediate-level short-lived radioactive waste will be built on the Stabatiškės site in the vicinity of Ignalina NPP during decommissioning works. The reservoir can also be used for the waste stored in the temporary repositories of the Ignalina NPP. Engineering and nature protective barriers are used in the repository for radioactive waste, however, radionuclides can spread into the environment, extend in the biosphere and cause (define the external power light exposure of the environment due to the natural and premature (prescheduled degradation of the engineering barriers of the repository. The properties of the soil (acidity, quantity of organic substances, humidity are being investigated for estimating the possible migration and dispersion of radionuclides. The activity of radionuclides in the soil is also estimated before building the repository. Natural and artificial radionuclides make the pollution of the soil, and therefore the accumulation and vertical migration of artificial (137Cs, 60Co and natural (226Ra, 232Th, 40K radionuclides are being researched in the soil on the Stabatiškės site.Article in Lithuanian

  11. NPP atucha I. 40 years of commercial operation of the heavy-water reactor in Argentina

    Mazzantini, Oscar A. [Nuclearelectrica Argentina SA, Atucha (Argentina); Fabian, Hermann O.

    2014-08-15

    The nuclear power plant (NPP) Atucha I in Argentina - a heavy-water reactor with pressure vessel technology operated with natural uranium - accomplished a remarkable anniversary on 26 June 2014: 40 years of commercial operation. State-run Nucleoelectrica Argentina SA (NA SA), being today the plant owner and operator commemorated this anniversary, as only few NPP exist which can refer to such long operating time with good performance. With a limited operating licence to 40 years (or rather 32 full load years) by the National Atomic Energy Agency (Comision National de Energia Atomica/CNEA) the plant had been handed over to CNEA on 24 June 1974 by the general contractor, Siemens AG, after release of the works contract on 1. June 1968. The site is located to the north-west of Buenos Aires upstream on the Rio Parana. The plant has an output of 345 MW; it has been continuously, reliable and successfully operated. Atucha I supplied overall 82.4 TWh of electricity into the national grid (220 kV) with an integral operating availability of 76.5 %.

  12. Reactor Dosimetry Aspects of the Service Life Extension of the Hungarian Paks NPP

    Zsolnay Eva M.

    2016-01-01

    Full Text Available The service life of the Hungarian Paks Nuclear Power Plant (NPP will be extended from the originally planned 30 years to 50 years. To improve the reliability of the results obtained in frame of the old reactor pressure vessel (RPV surveillance programme, new methods have been developed, and based on them, the old exposition data have been re-evaluated for all the four reactor units. At the same time, a new RPV surveillance programme has been developed and introduced, and long term irradiations have been performed to determine the radiation damage of the surveillance specimens due to the high fast neutron exposition. Neutron transport calculations have been performed with a validated neutron transport code system to determine the fast neutron exposition of the RPVs during the extended service life. The cavity dosimetry is in the introductory phase. This paper presents the new developments in the field of the RPV surveillance dosimetry and summarises the results obtained. According to the results the service life of the NPP can safely be extended for the planned 50 years.

  13. RADIATION CONDITIONS IN KALUGA REGION 30 YEARS AFTER CHERNOBYL NPP ACCIDENT

    A. G. Ashitko

    2016-01-01

    Full Text Available The article describes radiation conditions in the Kaluga region 30 years after the Chernobyl NPP accident. The Chernobyl NPP accident caused radioactive contamination of nine Kaluga region territories: Duminichsky, Zhizdrinsky, Kuibyshevsky, Kirovsky, Kozelsky, Ludinovsky, Meshchovsky, Ulyanovsky and Hvastovichsky districts. Radioactive fallout was the strongest in three southern districts: Zhizdrinsky, Ulyanovsky and Hvastovichsky, over there cesium-137 contamination density is from 1 to 15Ci/km. According to the Russian Federation Government Order in 2015 there are 300 settlements (S in the radioactive contamination zone, including 14 settlements with caesium-137 soil contamination density from 5 to 15 Ci/ km2 and 286 settlements with the contamination density ranging from 1 to 5 Ci/km2. In the first years after the Chernobyl NPP accident in Kaluga region territories, contaminated with caesium-137, there were introduced restrictive land usage, were carried out agrochemical activities (ploughing, mineral fertilizer dressing, there was toughened laboratory radiation control over the main doze-forming foodstuff. All these measures facilitated considerable decrease of caesium-137 content in local agricultural produce. Proceeding from the achieved result, in 2002 there took place the transition to more tough requirements SanPiN 2.3.2.1078-01. Analysis of investigated samples from Zhizdrinsky, Ulyanovsky and Hvastovichsky districts demonstrated that since 2005 meat samples didn’t exceed the standard values, same for milk samples since 2007. Till the present time, the use of wild-growing mushrooms, berries and wild animals meat involves radiation issues. It was demonstrated that average specific activity of caesium-137 in milk samples keeps decreasing year after year. Long after the Chernobyl NPP accident, the main products forming internal irradiation doses in population are the wild-growing mushrooms and berries. Population average annual

  14. Modeling the critical safety functions status tree of a NPP using FPGA

    Farias, Marcos Santana; Oliveira, Mauro Vitor de; Jaime, Guilherme Dutra Gonzaga; Almeida, Jose Carlos Soares de; Augusto, Silas Cordeiro, E-mail: msantana@ien.gov.br, E-mail: mvitor@ien.gov.br, E-mail: gdjaime@ien.gov.br, E-mail: jcsa@ien.gov.br, E-mail: silas@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Divisao de Engenharia Nuclear

    2013-07-01

    Field Programmable Gate Arrays (FPGAs) based systems and equipment are beginning to appear in new plants I and C applications, as well as in retrofits for operating plants, in particular for safety applications due to their capability to face the systems obsolescence since they are circuit independent. The circuits implemented can be portable to different FPGAs architectures. Moreover, they reduce complexity for regulatory approval as compared to conventional microprocessor-based systems. Critical safety function (CSF) is the most significant design concept for prioritize operator actions for NPP based on the potential threat to the three barriers (fuel cladding, primary coolant system boundary, and containment) and allows the operator to respond to these threats prior to event diagnosis. CSF has a hierarchical information structure that organizes the system variables affecting the plant safety in terms of goal-means relations. This paper describes the application of FPGA in the implementation of the CSFs status tree logic for a Westinghouse 3-loops NPP simulator. (author)

  15. Information technologies in radioactive waste management, applied in NPP-Kozloduy, Bulgaria

    Jeliazkov, J.; Jeliazkova, L.; Atanasov, Sv. [BALBOK Co., Sofia (Bulgaria)

    1994-12-31

    Radioactive waste (RAW) management in NPP is a complex problem, that can be considered as a combination of sub problems, for example scientific, administrative, social, economical, etc. The separate examination of these sub problems does not lead to creation of overall system for optimal RAW management. There`s no doubt that such an administrative system, supported by information technologies, should present in every one existing and planned nuclear power plant to optimize its operation as a whole, not only separate elements. The aim is to avoid the fallacy of the single and the complex. This paper presents a basic part of the whole information management system as defined above that concerns RAW management. The information management system is prepared for NPP-Kozloduy, Bulgaria by means of modern concepts and technological schemes and is aimed to help the administrative personnel in this very important activity - RAW management. On the base of objective data about the available waste and prognoses about arisings in the future, on the base of chosen technologies and equipment the system gives multi-variant plan for treatment, processing and disposal of waste, after the choice of a variant it monitors its application in the practice.

  16. Segmentation of the internal of the Reactor Jose Cabrera NPP; Segmentacion de los Internos del Reactor CN Jose Cabrera

    Rodriguez silva, M.; Borque Linan, J.

    2013-07-01

    The Plan of dismantling and decommissioning of the Jose Cabrera NPP represents the first total dismantling of a nuclear power station in Spain (level 3 of the IAEA). Complete disassembly of the different components of the primary circuit (internal reactor vessel, pusher, generator of steam, etc.) represents a differential activity against previous projects of dismantling The segmentation of the inmates of the reactor under water using tele operators cutting tools in the spent fuel pit, has been a challenge from the technological point of view as well as a critical activity in the framework of the radiological dis-assemblies associated to the Plan of dismantling and Decommissioning of the Jose Cabrera NPP.

  17. Methodology and measures for preventing unacceptable flow-accelerated corrosion thinning of pipelines and equipment of NPP power generating units

    Tomarov, G. V.; Shipkov, A. A.; Lovchev, V. N.; Gutsev, D. F.

    2016-10-01

    Problems of metal flow-accelerated corrosion (FAC) in the pipelines and equipment of the condensate- feeding and wet-steam paths of NPP power-generating units (PGU) are examined. Goals, objectives, and main principles of the methodology for the implementation of an integrated program of AO Concern Rosenergoatom for the prevention of unacceptable FAC thinning and for increasing operational flow-accelerated corrosion resistance of NPP EaP are worded (further the Program). A role is determined and potentialities are shown for the use of Russian software packages in the evaluation and prediction of FAC rate upon solving practical problems for the timely detection of unacceptable FAC thinning in the elements of pipelines and equipment (EaP) of the secondary circuit of NPP PGU. Information is given concerning the structure, properties, and functions of the software systems for plant personnel support in the monitoring and planning of the inservice inspection of FAC thinning elements of pipelines and equipment of the secondary circuit of NPP PGUs, which are created and implemented at some Russian NPPs equipped with VVER-1000, VVER-440, and BN-600 reactors. It is noted that one of the most important practical results of software packages for supporting NPP personnel concerning the issue of flow-accelerated corrosion consists in revealing elements under a hazard of intense local FAC thinning. Examples are given for successful practice at some Russian NPP concerning the use of software systems for supporting the personnel in early detection of secondary-circuit pipeline elements with FAC thinning close to an unacceptable level. Intermediate results of working on the Program are presented and new tasks set in 2012 as a part of the updated program are denoted. The prospects of the developed methods and tools in the scope of the Program measures at the stages of design and construction of NPP PGU are discussed. The main directions of the work on solving the problems of flow

  18. Sludge lancing and IBL: Results and experiences in the Spanish NPP; SLUDGE LANCING e IBL: Resultados y experiencias en las central espanolas

    Montoro, E.; Pozo, C. del

    2013-07-01

    During the cycle of operation of the PWR plants, oxides (sludge) tanks are generated in the secondary circuit by corrosion, chemical additives, etc which are deposited onto the tubular steam generators (GVs), limiting its efficiency and service life.

  19. Climate change impacts on net primary production (NPP) and export production (EP) regulated by increasing stratification and phytoplankton community structure in the CMIP5 models

    Fu, Weiwei; Randerson, James T.; Moore, J. Keith

    2016-09-01

    We examine climate change impacts on net primary production (NPP) and export production (sinking particulate flux; EP) with simulations from nine Earth system models (ESMs) performed in the framework of the fifth phase of the Coupled Model Intercomparison Project (CMIP5). Global NPP and EP are reduced by the end of the century for the intense warming scenario of Representative Concentration Pathway (RCP) 8.5. Relative to the 1990s, NPP in the 2090s is reduced by 2-16 % and EP by 7-18 %. The models with the largest increases in stratification (and largest relative declines in NPP and EP) also show the largest positive biases in stratification for the contemporary period, suggesting overestimation of climate change impacts on NPP and EP. All of the CMIP5 models show an increase in stratification in response to surface-ocean warming and freshening, which is accompanied by decreases in surface nutrients, NPP and EP. There is considerable variability across the models in the magnitudes of NPP, EP, surface nutrient concentrations and their perturbations by climate change. The negative response of NPP and EP to increasing stratification reflects primarily a bottom-up control, as upward nutrient flux declines at the global scale. Models with dynamic phytoplankton community structure show larger declines in EP than in NPP. This pattern is driven by phytoplankton community composition shifts, with reductions in productivity by large phytoplankton as smaller phytoplankton (which export less efficiently) are favored under the increasing nutrient stress. Thus, the projections of the NPP response to climate change are critically dependent on the simulated phytoplankton community structure, the efficiency of the biological pump and the resulting levels of regenerated production, which vary widely across the models. Community structure is represented simply in the CMIP5 models, and should be expanded to better capture the spatial patterns and climate-driven changes in export

  20. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  1. Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

    Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-01-01

    This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10{sup 4} to 3.85 x 10{sup 5}. (Author) 5 refs., 18 figs., 5 tabs.

  2. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.

    1978-09-01

    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

  3. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  4. PWR type reactors. Normal and accidental operation; Reacteurs a eau sous pression. Fonctionnement normal et accidentel

    Petetrot, J.F. [AREVA NP, Dept. Fonctionnement Reacteur et Etudes d' Accidents/Division, Tour AREVA, 92 - Paris La Defense (France)

    2009-07-15

    This article presents the general operation principles of PWR type reactors with the limits to be respected for the core and the steam supply system. Regulation systems controlling the main parameters are described as well: measurements used, functional structures, controlled actuators. The protection system which can lead to the automatic shutdown of the reactor (emergency rod drop) and to the start-up of safeguard functions is detailed. The interface for the conventional protection system is briefly described. The operation of the steam supply system with respect to the power grid needs is presented in relation with the regulation of the turbogenerator set: load follow-up, primary and secondary adjustment. Finally, the changes of the most important parameters during typical transients are commented. The main operations needed to move from the cold shutdown state to the nominal power are described as well. (J.S.)

  5. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  6. Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference

    Castro, Emilio; Buss, Oliver; Garcia-Herranz, Nuria; Hoefer, Axel; Porsch, Dieter

    2016-01-01

    A Monte Carlo-based Bayesian inference model is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous...

  7. Numerical modeling of in-vessel melt water interaction in large scale PWR`s

    Kolev, N.I. [Siemens AG, KWU NA-M, Erlangen (Germany)

    1998-01-01

    This paper presents a comparison between IVA4 simulations and FARO L14, L20 experiments. Both experiments were performed with the same geometry but under different initial pressures, 51 and 20 bar respectively. A pretest prediction for test L21 which is intended to be performed under an initial pressure of 5 bar is also presented. The strong effect of the volume expansion of the evaporating water at low pressure is demonstrated. An in-vessel simulation for a 1500 MW el. PWR is presented. The insight gained from this study is: that at no time are conditions for the feared large scale melt-water intermixing at low pressure in force, with this due to the limiting effect of the expansion process which accelerates the melt and the water into all available flow paths. (author)

  8. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  9. Transient fuel behavior of preirradiated PWR fuels under reactivity initiated accident conditions

    Fujishiro, Toshio; Yanagisawa, Kazuaki; Ishijima, Kiyomi; Shiba, Koreyuki

    1992-06-01

    Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

  10. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail: rochkhudson@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear

    2015-07-01

    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  11. Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

    Taniguchi, T.; Ligon, D.; Stamatelatos, M.

    1983-04-01

    Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights in the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.

  12. PRETTA:A COMPUTER PROGRAM FOR PWR PRESSURIZER’S TRANSIENT THERMODYNAMICS

    阿谢德; 徐济鋆

    2001-01-01

    A computer program PRETTA “Pressurizer Transient Thermodynamics Analysis” was developed for the prediction of pressurizer under transient conditions. It is based on the solution of the conservation laws of heat and mass applied to the three separate and non equilibrium thermodynamic regions. In the program all of the important thermal-hydraulics phenomena occurring in the pressurizer: stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer have been considered. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented computer program will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant's pressurizer performance.

  13. Effect of Weld Properties on the Crush Strength of the PWR Spacer Grid

    Kee-nam Song

    2012-01-01

    Full Text Available Mechanical properties in a weld zone are different from those in the base material because of different microstructures. A spacer grid in PWR fuel is a structural component with an interconnected and welded array of slotted grid straps. Previous research on the strength analyses of the spacer grid was performed using base material properties owing to a lack of mechanical properties in the weld zone. In this study, based on the mechanical properties in the weld zone of the spacer grid recently obtained by an instrumented indentation technique, the strength analyses considering the mechanical properties in the weld zone were performed, and the analysis results were compared with the previous research.

  14. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    Hermann, O.W.

    1999-09-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

  15. Representing Operational Knowledge of PWR Plant by Using Multilevel Flow Modelling

    Zhang, Xinxin; Lind, Morten; Jørgensen, Sten Bay;

    2014-01-01

    situation and support operational decisions. This paper will provide a general MFM model of the primary side in a standard Westinghouse Pressurized Water Reactor ( PWR ) system including sub - systems of Reactor Coolant System, Rod Control System, Chemical and Volume Control System, emergency heat removal......The aim of this paper is to explore the capability of representing operational knowledge by using Multilevel Flow Modelling ( MFM ) methodology. The paper demonstrate s how the operational knowledge can be inserted into the MFM models and be used to evaluate the plant state, identify the current...... systems. And the sub - systems’ functions will be decomposed into sub - models according to different operational situations. An operational model will be developed based on the operating procedure by using MFM symbols and this model can be used to implement coordination rules for organize the utilizati...

  16. Steady characteristic investigation on passive residual heat removal system of Chinese advanced PWR

    2008-01-01

    Thermal-hydraulic characteristic investigation on passive residual heat removal system(PRHRS)of Chinese advanced PWR was conducted to provide input data for PRHRS design and to demonstrate the feasibility of unique design features.A total of 237 sets of test data at steady state have been obtained and the main influence factors on the two-phase natural circulation flow rate and residual heat removal capability were identified.On the basis of theory analysis,a correlation of two-phase natural circulation was obtained,and relative errors of 95% test data were less than±16%.There is a considerable effect of the system status parameters on the threshold of height between heat source and heat sink,and its correlation of two-phase natural circulation system has been obtained.The steady characteristic research shows that PRHRS has the capability of removing the core decay power through natural circulation.

  17. Fatigue Crack Growth Rate of Type 347 Stainless Steel at the PWR Environment

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Materials used in nuclear power plants are low alloy steel, stainless steel, and superalloy steel. Understanding the characteristics of these materials is important in the development of nuclear power plant related technology. Nb-stabilized Type 347 stainless steel is used for the coolant pressurizer surge line of Korea Standard Nuclear Power Plant (KSNPP). Surge line of PWR nuclear reactor are damaged by thermal fatigue due to thermal gradient during heat-up and cool-down, mechanical fatigue due to mechanical stress, and corrosion fatigue due to nuclear reactor water environment. Fatigue is an important factor which limits the life of structure. Fatigue crack growth rate curves in nuclear reactor environment are needed to evaluate the integrity of nuclear reactor structure but that result is not sufficient. In this study, fatigue crack growth rates at nuclear reactor environment are produced to evaluate integrity of nuclear power plant section 5

  18. The radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR

    Kelly, G N; Charles, D; Hemming, C R

    1983-01-01

    This report contains an assessment of the radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR. Three of the degraded core accident releases postulated by the CEGB are analysed. The consequences, conditional upon each release, are evaluated in terms of the health impact on the exposed population and the impact of countermeasures taken to limit the exposure. Consideration is given to the risk to the Greater London population as a whole and to individuals within it. The consequences are evaluated using the NRPB code MARC (Methodology for Assessing Radiological Consequences). The results presented in this report are all conditional upon the occurrence of each release. In assessing the significance of the results, due account must be taken of the frequency with which such releases may be predicted to occur.

  19. Analysis of nuclear characteristics and fuel economics for PWR core with homogeneous thorium fuels

    Joo, H. K.; Noh, J. M.; Yoo, J. W.; Song, J. S.; Kim, J. C.; Noh, T. W

    2000-12-01

    The nuclear core characteristics and economics of an once-through homogenized thorium cycle for PWR were analyzed. The lattice code, HELIOS has been qualified against BNL and B and W critical experiments and the IAEA numerical benchmark problem in advance of the core analysis. The infinite multiplication factor and the evolution of main isotopes with fuel burnup were investigated for the assessment of depletion charateristics of thorium fuel. The reactivity of thorium fuel at the beginning of irradiation is smaller than that of uranium fuel having the same inventory of {sup 235}U, but it decrease with burnup more slowly than in UO{sub 2} fuel. The gadolinia worth in thorium fuel assembly is also slightly smaller than in UO{sub 2} fuel. The inventory of {sup 233}U which is converted from {sup 232}Th is proportional to the initial mass of {sup 232}Th and is about 13kg per one tones of initial heavy metal mass. The followings are observed for thorium fuel cycle compared with UO{sub 2} cycle ; shorter cycle length, more positive MTC at EOC, more negative FTC, similar boron worth and control rod. Fuel economics of thorium cycle was analyzed by investigating the natural uranium requirements, the separative work requirements, and the cost for burnable poison rods. Even though less number of burnable poison rods are required in thorium fuel cycle, the costs for the natural uranium requirements and the separative work requirements are increased in thorium fuel cycle. So within the scope of this study, once through cycle concept, homogenized fuel concept, the same fuel management scheme as uranium cycle, the thorium fuel cycle for PWR does not have any economic incentives in preference to uranium.

  20. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  1. Singular deposit formation in PWR due to electrokinetic phenomena - application to SG clogging

    Guillodo, M.; Muller, T.; Barale, M.; Foucault, M. [AREVA NP SAS, Technical Centre (France); Clinard, M.-H.; Brun, C.; Chahma, F. [AREVA NP SAS, Chemistry and Radiochemistry Group (France); Corredera, G.; De Bouvier, O. [Electricite de France, Centre d' Expertise de I' inspection dans les domaines de la Realisation et de l' Exploitation (France)

    2009-07-01

    The deposits which cause clogging of the 'foils' of the tube support plates (TSP) in Steam Generators (SG) of PWR present two characteristics which put forward that the mechanism at the origin of their formation is different from the mechanism that drives the formation of homogeneous deposits leading to the fouling of the free spans of SG tubes. Clogging occurs near the leading edge of the TSP and the deposits appear as diaphragms localized between both TSP and SG tubing materials, while the major part of the tube/TSP interstice presents little or no significant clogging. This type of deposit seems rather comparable to the ones which were reproduced in Lab tests to explain the flow rate instabilities observed on a French unit during hot shutdown in the 90's. The deposits which cause TSP clogging are owed to a discontinuity of the streaming currents in the vicinity of a surface singularity (orifices, scratches ...) which, in very low conductivity environment, produce local potential variations and/or current loop in the metallic pipe material due to electrokinetic effects. Deposits can be built by two mechanisms which may or not coexist: (i) accumulation of particles stabilized by an electrostatic attraction due to the local variation of electrokinetic potential, and (ii) crystalline growth of magnetite produced by the oxidation of ferrous ions on the anodic branch of a current loop. Lab investigations carried out by AREVA NP Technical Centre since the end of the 90's showed that this type of deposit occurs when the redox potential is higher than a critical value, and can be gradually dissolved when the potential becomes lower than this value which depends on the 'Material - Chemistry' couple. Special emphasis will be given in this paper to the TSP clogging of SG in PWR secondary coolant dealing particularly with the potential strong effect of electrokinetic phenomena in low conductive environment and in high temperature conditions

  2. An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel

    DeHart, M.D.

    1993-01-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms for subsequent system analyses involving heat transfer, radiation shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. Such comparisons have been performed in earlier work at the Oak Ridge National Laboratory (ORNL). This report describes additional independent measurements and corresponding calculations, which supplement the results of the earlier work. The current work includes measured isotopic data from 19 spent fuel samples obtained from the Italian Trino Vercelles pressurized-water reactor (PWR) and the U.S. Turkey Point Unit 3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results are presented based on the combination of measured-to-calculated ratios for earlier work and the current analyses. The results described herein represent an extension to a new reactor design not included in the earlier work, and spent fuel samples with enrichment as high as 3.9 wt % {sup 235}U. Results for the current work are found to be, for the most part, consistent with the findings of the earlier work. This consistency was observed for results obtained from each of two different cross-section libraries and suggests that the estimated biases determined for

  3. Land and Cryosphere Products from Suomi NPP VIIRS: Overview and Status

    Roman, M. O.; Justice, C. O.; Csiszar, I. A.; Vermote, E.; Wolfe, R. E.; Hook, S. J.; Friedl, M. A.; Schaaf, C.; Wang, Z.; Miura, T.; Tschudi, M. A.; Riggs, G. A.; Hall, D. K.; Lyapustin, A.; Devadiga, S.; Davidson, C.; Masuoka, E. J.

    2013-12-01

    The Visible Infrared Imaging Radiometer Suite (VIIRS) instrument was launched in October 2011 as part of the Suomi National Polar-orbiting Partnership (S-NPP). The VIIRS instrument was designed to improve upon the capabilities of the operational Advanced Very High Resolution Radiometer (AVHRR) and provide observation continuity with NASA's Earth Observing System's (EOS) Moderate Resolution Imaging Spectroradiometer (MODIS). Since the VIIRS first-light images were received in November 2011, NASA and NOAA funded scientists have been working to evaluate the instrument performance and generate land and cryosphere products which would meet the needs of the NOAA operational users and the NASA science community. NOAA's focus has been on refining a suite of operational products known as Environmental Data Records (EDRs), which were developed according to project specifications under the National Polar-orbiting Environmental Satellite System (NPOESS). The NASA S-NPP Science Team has focused on evaluating the EDRs for science use and developing and testing additional products to meet science data needs and provide MODIS data product continuity. This paper will present the findings to-date of the NASA Science Team's evaluation of the VIIRS Land and Cryosphere EDRs, specifically Surface Reflectance, Land Surface Temperature, Surface Albedo, Vegetation Indices, Surface Type, Active Fires, Snow Cover, Ice Surface Temperature, and Sea Ice Characterization. To achieve the stated goal of MODIS data continuity and the establishment of long-term data records through VIIRS, it is important to start now to use S-NPP to establish a pathway to science use of VIIRS data in the JPSS era. One year after launch, initial instrument and operational product evaluations are now ending and the next step is to build on the success of the MODIS Adaptive Processing System (MODAPS) and the Land Product Evaluation and Analysis Tool Element (PEATE) data processing and generate and distribute high

  4. A Cloud-Based Infrastructure for Near-Real-Time Processing and Dissemination of NPP Data

    Evans, J. D.; Valente, E. G.; Chettri, S. S.

    2011-12-01

    We are building a scalable cloud-based infrastructure for generating and disseminating near-real-time data products from a variety of geospatial and meteorological data sources, including the new National Polar-Orbiting Environmental Satellite System (NPOESS) Preparatory Project (NPP). Our approach relies on linking Direct Broadcast and other data streams to a suite of scientific algorithms coordinated by NASA's International Polar-Orbiter Processing Package (IPOPP). The resulting data products are directly accessible to a wide variety of end-user applications, via industry-standard protocols such as OGC Web Services, Unidata Local Data Manager, or OPeNDAP, using open source software components. The processing chain employs on-demand computing resources from Amazon.com's Elastic Compute Cloud and NASA's Nebula cloud services. Our current prototype targets short-term weather forecasting, in collaboration with NASA's Short-term Prediction Research and Transition (SPoRT) program and the National Weather Service. Direct Broadcast is especially crucial for NPP, whose current ground segment is unlikely to deliver data quickly enough for short-term weather forecasters and other near-real-time users. Direct Broadcast also allows full local control over data handling, from the receiving antenna to end-user applications: this provides opportunities to streamline processes for data ingest, processing, and dissemination, and thus to make interpreted data products (Environmental Data Records) available to practitioners within minutes of data capture at the sensor. Cloud computing lets us grow and shrink computing resources to meet large and rapid fluctuations in data availability (twice daily for polar orbiters) - and similarly large fluctuations in demand from our target (near-real-time) users. This offers a compelling business case for cloud computing: the processing or dissemination systems can grow arbitrarily large to sustain near-real time data access despite surges in

  5. Selection of optimal treatment procedures for non-standard radioactive waste arising from decommissioning of NPP after accident

    Strážovec, Roman, E-mail: strazovec.roman@javys.sk [Institute of Nuclear and Physical Engineering, Slovak University of Technology in Bratislava, Ilkovičova 3, 812 19 Bratislava (Slovakia); JAVYS, a.s., Tomášikova 22, 821 02 Bratislava (Slovakia); Hrnčíř, Tomáš [DECOM, a.s., Sibírska 1, 917 01 Trnava (Slovakia); Lištjak, Martin [Institute of Nuclear and Physical Engineering, Slovak University of Technology in Bratislava, Ilkovičova 3, 812 19 Bratislava (Slovakia); VUJE, a.s., Okružná 5, 918 64 Trnava (Slovakia); Nečas, Vladimír [Institute of Nuclear and Physical Engineering, Slovak University of Technology in Bratislava, Ilkovičova 3, 812 19 Bratislava (Slovakia)

    2016-05-15

    The decommissioning of nuclear power plants is becoming a standard industrial activity where the optimization processes of partial activities are inevitable mainly for technical and economic reasons. In Slovakia, the decommissioning of A1 NPP is very specific case because A1 NPP is rare type of NPP (prototype) and furthermore its operation was affected by the accident. A large number of specific non-standard radioactive waste, such as long-time storage cases (hereinafter LSC), that is not usually present within the decommissioning projects of NPP with a regular termination of operation, represent one of the significant consequences of the accident and issues arisen from follow-up activities. The presented article describes the proposal of processing and conditioning of non-standard radioactive waste (such as LSC), together with description of methodology applied in the proposal for update of waste acceptance criteria for the processing and conditioning of radioactive waste (hereinafter RAW) within Bohunice Radioactive waste Treatment and Conditioning Centre (hereinafter RWTC). The results of performed detailed analysis are summarized into new waste acceptance criteria for technological lines keeping in mind safety principles and requirements for protection of operating personnel, the public and the environment.

  6. Near-surface final repository for the NPP Chernobyl, Ukraine; Oberflaechennahes Endlager fuer das KKW Tschernobyl, Ukraine

    Eichhorn, Heiko [NUKEM Technologies GmbH (Germany)

    2008-07-01

    NUKEM Technologies has realized the project of near-surface final radioactive waste storage at the site of the NPP Chernobyl. The complex includes facilities for treatment and processing of solid wastes, interim storage facilities for high-level waste and a near-surface final repository for conditioned radioactive waste. The project is funded by the European Union in the frame of TACIS.

  7. 76 FR 46330 - NUREG-1934, Nuclear Power Plant Fire Modeling Application Guide (NPP FIRE MAG); Second Draft...

    2011-08-02

    ... COMMISSION NUREG-1934, Nuclear Power Plant Fire Modeling Application Guide (NPP FIRE MAG); Second Draft... for public comment a document entitled, NUREG-1934 (EPRI 1023259), ``Nuclear Power Plant Fire Modeling... pdr.resource@nrc.gov . NUREG-1934 (EPRI 1023259), ``Nuclear Power Plant Fire Modeling...

  8. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  9. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

    2014-07-01

    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  10. Chemical mode in secondary circuit of the Dukovany NPP units after TG condensers replacement

    Kopriva, M.; Shejbal, J.; Petrecky, I. [Dukovany NPP (Czech Republic)

    2002-07-01

    The increase of the pH of SG feedwater on the 1. unit of Dukovany NPP led to enhancement of chemical mode of secondary circuit, what was identified in particular by the following: Reduction of concentration of iron in SG feedwater, Reduction of concentration of Sodium and Sulfates in SG blowdown water. This reduction is caused by shutdown of CPS thus by elimination of release of Na ions and SO{sub 4} from wrong operated ion-exchangers and their subsequent regeneration (part of cation exchanger in Na form and part of anion exchanger in SO{sub 4} form). Reduction of the WANO SG chemical index to the minimum theoretical value. It will be necessary to change criteria characterizing this index or to introduce our own modified index. In relation to CPS shutdown the costs for operating chemicals and for demineralized flushing water were reduced. (authors)

  11. ELECTRICALLY CONDUCTIVE OF NANOCOMPOSITES FOR SYSTEMS DIAGNOSTICS OF THE ENVELOPE WALLS TECHNICAL CONDITION OF NPP

    BOLSHAKOV V. I.

    2016-05-01

    Full Text Available Raising of the problem. Enveloped concrete wall type structures of localizing safety systems for restaint and localization of radioactive decay products or in the case of special natural or man-made impacts on the power unit is one of the most important components to ensure the safety of nuclear power. The promising direction for the development of the NPP technical system monitoring is to use conductive nanocomposites as primary elements of information. The purpose of the article is to review the theoretical background and experience in the conductive nanocomposites creating for diagnostics of localizing nuclear safety systems. Conclusions. A promising area for the development of diagnostic systems of localizing nuclear safety systems is the use of electrically conductive nanocomposites (conductive concrete - bethels, plasters, paint coatings. A mechanism for conductive nanocomposites creating is the use of the filler metal and carbon nanoparticles. As binders is promising to use nanocomposites of the mineral binders (cement and water glass.

  12. Surface pathway of radioactive plume of TEPCO Fukushima NPP1 released 134Cs and 137Cs

    M. Aoyama

    2013-05-01

    Full Text Available 134Cs and 137Cs were released to the North Pacific Ocean by two major likely pathways, direct discharge from the Fukushima NPP1 accident site and atmospheric deposition off Honshu Islands of Japan, east and northeast of the site. High density observations of 134Cs and 137Cs in the surface water were carried out by 17 cruises of cargo ships and several research vessel cruises from March 2011 till March 2012. The main body of radioactive surface plume of which activity exceeded 10 Bq m−3 travelled along 40° N and reached the International Date Line on March 2012, one year after the accident. A distinct feature of the radioactive plume was that it stayed confined along 40° N when the plume reached the International Date Line. A zonal speed of the radioactive plume was estimated to be about 8 cm s−1 which was consistent with zonal speeds derived by Argo floats at the region.

  13. Photovoltaic characteristics of n(+)pp(+) InP solar cells grown by OMVPE

    Tyagi, S.; Singh, K.; Bhimnathwala, H.; Ghandhi, S. K.; Borrego, J. M.

    1990-01-01

    The photovoltaic characteristics of n(+)/p/p(+) homojunction InP solar cells fabricated by organometallic vapor-phase epitaxy (OMVPE) are described. The cells are characterized by I-V, C-V and quantum efficiency measurements, and simulations are used to obtain various device and material parameters. The I-V characteristics show a high recombination rate in the depletion region; this is shown to be independent of the impurity used. It is shown that cadmium is easier to use as an acceptor for the p base and p(+) buffer and is therefore beneficial. The high quantum efficiency of 98 percent at long wavelengths measured in these cells indicates a very good collection efficiency in the base. The short-wavelength quantum efficiency is poor, indicating a high surface recombination.

  14. Sulfate and Chloride Resistance of High Fluidity Concrete including Fly Ash and GGBS for NPP

    Noh, Jea Myoung; Cho, Myung Sug [KEPCO Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Fly ash mixed concrete has been used for NPP concrete structures in Korea in order to prevent aging and improve durability since the Shin.Kori no.1,2 in 2005. Concentrated efforts to develop technology for the streamlining of construction work and to affect labor savings have been conducted in construction. The application of high fluidity concrete for nuclear power plants has been the research subject with the aim of further rationalization of construction works. Since high fluidity concrete can have the characteristics of high density and high strength without compaction. However, high fluidity concrete can cause thermal cracking by heat of hydration. For this reason, the amount of pozzolan binder should be increased in high fluidity concrete for nuclear power plants. In this study, the resistance of high fluidity concrete on sulfate and chloride was compared with that of the concrete currently using for nuclear power plants

  15. Developing Land Surface Type Map with Biome Classification Scheme Using Suomi NPP/JPSS VIIRS Data

    Zhang, Rui; Huang, Chengquan; Zhan, Xiwu; Jin, Huiran

    2016-08-01

    Accurate representation of actual terrestrial surface types at regional to global scales is an important element for a wide range of applications, such as land surface parameterization, modeling of biogeochemical cycles, and carbon cycle studies. In this study, in order to meet the requirement of the retrieval of global leaf area index (LAI) and fraction of photosynthetically active radiation absorbed by the vegetation (fPAR) and other studies, a global map generated from Suomi National Polar- orbiting Partnership (S-NPP) Visible Infrared Imaging Radiometer Suite (VIIRS) surface reflectance data in six major biome classes based on their canopy structures, which include: Grass/Cereal Crops, Shrubs, Broadleaf Crops, Savannas, Broadleaf Forests, and Needleleaf Forests, was created. The primary biome classes were converted from an International Geosphere-Biosphere Program (IGBP) legend global surface type data that was created in previous study, and the separation of two crop types are based on a secondary classification.

  16. Study of Degradation Processes in Dielectric Materials Used in Electronic Control Equipment Operated in ``Kozloduy'' NPP

    Naydenov, Nayden; Popov, Angel

    2007-04-01

    The electronic equipment for control of different systems of Units 5 and 6 is studied for presence of degradation processes occurring in result of continuous usage in conditions of controlled radiation background in compliance with ``Kozloduy'' NPP safety codes. Systems, operated in a continuous mode in the course of about 10 years were chosen - separate units containing different dielectric materials (varnish coating, circuit board bases, cable insulations, electro protective elements, etc.) were extrapolated. Series of test samples were prepared which were connected with flat or coaxial condensers and their characteristic parameters were measured: tgδ, ɛ, low voltage conductivity and leak currents at voltages that exceed the working ones several times. When comparing the obtained data with the reference ones, a conclusion is made about the effectiveness of electric ageing during operation in the course of time.

  17. Development of Rainfall-Discharge Model for Future NPP candidate Site

    An, Ji-hong; Yee, Eric [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    By this study, most suitable model for future nuclear power plant site in Yeongdeok to be used to predict peak amount of riverine flooding was developed by examining historical rainfall and discharge data from the nearest gage station which is Jodong water level gage station in Taehwa basin. Sitting a nuclear power plant (NPP) requires safety analyses that include the effects of extreme events such as flooding or earthquake. In light of South Korean government's 15-year power supply plan that calls for the construction of new nuclear power station in Yeongdeok, it becomes more important to site new station in a safe area from flooding. Because flooding or flooding related accidents mostly happen due to extremely intense rainfall, it is necessary to find out the relationship between rainfall and run-off by setting up feasible model to figure out the peak flow of the river around nuclear related facilities.

  18. Retrospective Dosimetry of Vver 440 Reactor Pressure Vessel at the 3RD Unit of Dukovany Npp

    Marek, M.; Viererbl, L.; Sus, F.; Klupak, V.; Rataj, J.; Hogel, J.

    2009-08-01

    Reactor pressure vessel (RPV) residual lifetime of the Czech VVER-440 is currently monitored under Surveillance Specimens Programs (SSP) focused on reactor pressure vessel materials. Neutron fluence in the samples and its distribution in the RPV are determined by a combination of calculation results and the experimental data coming from the reactor dosimetry measurements both in the specimen containers and in the reactor cavity. The direct experimental assessment of the neutron flux density incident onto RPV and neutron fluence for the entire period of nuclear power plant unit operation can be based on the evaluation of the samples taken from the inner RPV cladding. The Retrospective Dosimetry was also used at Dukovany NPP at its 3rd unit after the 18th cycle. The paper describes methodology, experimental setup for sample extraction, measurement of activities, and the determination of the neutron flux and fluence averaged over the samples.

  19. NIRS external dose estimation system for Fukushima residents after the Fukushima Dai-ichi NPP accident

    Akahane, Keiichi; Yonai, Shunsuke; Fukuda, Shigekazu; Miyahara, Nobuyuki; Yasuda, Hiroshi; Iwaoka, Kazuki; Matsumoto, Masaki; Fukumura, Akifumi; Akashi, Makoto

    2013-04-01

    The great east Japan earthquake and subsequent tsunamis caused Fukushima Dai-ichi Nuclear Power Plant (NPP) accident. National Institute of Radiological Sciences (NIRS) developed the external dose estimation system for Fukushima residents. The system is being used in the Fukushima health management survey. The doses can be obtained by superimposing the behavior data of the residents on the dose rate maps. For grasping the doses, 18 evacuation patterns of the residents were assumed by considering the actual evacuation information before using the survey data. The doses of the residents from the deliberate evacuation area were relatively higher than those from the area within 20 km radius. The estimated doses varied from around 1 to 6 mSv for the residents evacuated from the representative places in the deliberate evacuation area. The maximum dose in 18 evacuation patterns was estimated to be 19 mSv.

  20. Modeling the dynamics of distribution, extent, and NPP of global terrestrial ecosystems in response to future climate change

    Gang, Chengcheng; Zhang, Yanzhen; Wang, Zhaoqi; Chen, Yizhao; Yang, Yue; Li, Jianlong; Cheng, Jimin; Qi, Jiaguo; Odeh, Inakwu

    2017-01-01

    Understanding how terrestrial ecosystems would respond to future climate change can substantially contribute to scientific evaluation of the interactions between vegetation and climate. To reveal the future climate impacts might on the nature and magnitude of global vegetation, the spatiotemporal distribution and net primary productivity (NPP) of global terrestrial biomes and their dynamics in this century were quantitatively simulated and compared by using the improved Comprehensive and Sequential Classification System and the segmentation model. The 33 general circulation models under the four scenarios of Representative Concentration Pathways (RCPs) were utilized to simulate the future climate change. The multi-model ensemble results showed that at the global scale, the distribution of forests and deserts would expand by more than 2% and 4% over this century, respectively. By contrast, more than 11% of grassland regions would shrink. Despite the considerable differences in the simulated responses of the biomes, the poleward movement or expansion of temperate forest were prominent features across all the scenarios. Meanwhile, the terrestrial NPP was projected to increase by 7.44, 9.51, 9.46, and 12.02 Pg DW·a- 1 in 2070s relative to 1970s in the RCP2.6, RCP4.5, RCP6.0, and RCP8.5, respectively. The largest NPP decrease would occur in tundra & alpine steppe. NPP in the Tropical Zone, the North Temperate Zone, and the North Frigid Zone was estimated to increase in this century, whereas NPP in the South Temperate Zone was projected to decrease slightly across all scenarios. Overall, ecosystems in the mid-/high latitudes would be more vulnerable to future climate change in terms of distribution ranges and primary productivity despite the existing uncertainties. Some vegetation would benefit from the warmer and wetter climate. However, most of these plants would suffer and experience irreversible changes, particularly in the northern hemisphere.

  1. Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC; Desarrollo de un modelo del NSSS de un reactor PWR con el codigo termo-hidraulico GOTHIC

    Gomez Garcia-Torano, I.; Jimenez, G.

    2013-07-01

    The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

  2. Operational diagnostics of thermal state and efficiency of steam turbines of TPP and NPP

    Bozhko, V. V.; Kovalenko, A. N.; Lyapunov, V. M.; Khomenok, L. A.

    2016-05-01

    Various ways for solving complex problems of the strength and operating life of steam turbines of TPP and NPP are studied. Diagnostic characters and technical possibilities for their control during the steam turbine operation are determined. It is shown that the effect of various factors on the reliability, maneuverability, and service life of power installations of TPP and NPP is generally determined by the thermal state of steam-electric generating sets. Leading foreign and domestic manufacturers give major attention to the organization of the control of the thermal state of facilities and the development of systems for accounting ("counter") the service life depletion. Zones of high-temperature sites of shafts and disks with maximum parameters of operating environment are determined. A model for on-line computation of thermal stresses with the diagnostic evaluation of the service life depletion (fatigue accumulation) and forecasting of optimum heating conditions for thermostressed turbine units is briefly stated. An example of a program for diagnostics of the quality of the facility operation is given. The program provides the operative control of thermal stresses and the service life depletion in main units of the turbine under various operation conditions, operates in the real-time mode, calculates and represents currents values of thermal stresses in turbine units, and forms and transmits into the industrial control signals on the occurrence of restrictions with respect to thermal stresses and prohibition of an increase or decrease in the vapor temperature and the load in the case of approaching pressures to maximum permissible ones. In the case of stationary operation conditions, the program computed the current efficiency in high (HPC) and mean (MPC) pressure cylinders.

  3. Patterns of NPP, GPP, Respiration and NEP During Boreal Forest Succession

    Goulden, Michael L.; McMillan, Andrew; Winston, Greg; Rocha, Adrian; Manies, Kristen; Harden, Jennifer W.; Bond-Lamberty, Benjamin

    2010-12-15

    We deployed a mesonet of year-round eddy covariance towers in boreal forest stands that last burned in ~1850, ~1930, 1964, 1981, 1989, 1998, and 2003 to understand how CO2 exchange changes during secondary succession.The strategy of using multiple methods, including biometry and micrometeorology, worked well. In particular, the three independent measures of NEP during succession gave similar results. A stratified and tiered approach to deploying eddy covariance systems that combines many lightweight and portable towers with a few permanent ones is likely to maximize the science return for a fixed investment. The existing conceptual models did a good job of capturing the dominant patterns of NPP, GPP, Respiration and NEP during succession. The initial loss of carbon following disturbance was neither as protracted nor large as predicted. This muted response reflects both the rapid regrowth of vegetation following fire and the prevalence of standing coarse woody debris following the fire, which is thought to decay slowly. In general, the patterns of forest recovery from disturbance should be expected to vary as a function of climate, ecosystem type and disturbance type. The NPP decline at the older stands appears related to increased Rauto rather than decreased GPP. The increase in Rauto in the older stands does not appear to be caused by accelerated maintenance respiration with increased biomass, and more likely involves increased allocation to fine root turnover, root metabolism, alternative forms of respiration, mycorrhizal relationships, or root exudates, possibly associated with progressive nutrient limitation. Several studies have now described a similar pattern of NEP following boreal fire, with 10-to-15 years of modest carbon loss followed by 50-to-100 years of modest carbon gain. This trend has been sufficiently replicated and evaluated using independent techniques that it can be used to quantify the likely effects of changes in boreal fire frequency and

  4. The Use of Tritiated Wastewater from NPP Cernavoda in Agigea Black Sea Area

    Varlam, C.; Stefanescu, I.; Faurescu, I.; Vagner, I.; Dobrinescu, D.

    2009-04-01

    Danube-Black Sea Channel is situated in the south east part of Romania. It takes its waters from Danube upstream of Cernavoda town, and flows into the Black Sea at Agigea. The main uses of the channel are numerous, but it can be mentioned navigation and nuclear power generation - Cernavoda Nuclear Power Plant. Maximum weigh carrying of the canal is 70 million t/year. Agigea is the most important harbor of the Channel and due to intensive activity the accidents can happened any time. In this study we propose to use tritiated liquid effluents from CANDU type NPP Cernavoda as a tracer, to study dilution factor between fresh water of the canal and the Black Sea waters. Tritiated water can be used to simulate the transport and dispersion of solutes in Danube-Black Sea Channel because they have the same physical characteristics as water. Measured tracer-response curves produced from injection of a known quantity of soluble tracer provide an efficient method of obtaining necessary data. We established tritium level in monitored zone Agigea-Black Sea by monthly samples and precipitation during may 2005- may 2006. The average tritium concentration for Black Sea near Agigea location was 12.5 +/- 2.2 TU. During the NPP evacuation we establish inside Agigea canal lock a well mixed tritium concentration of 174.07 +/- 6.2 TU. Using waters lock as a tracer we establish dilution factor of 1: 4.28 inside Agigea harbor, and dilution factor of 1:2.17 in open sea. This experimental parameter can be used in dispersion simulation for Agigea - Black Sea area.

  5. Failure Forewarning in NPP Equipment NERI2000-109 Final Project Report

    Hively, LM

    2004-03-26

    The objective of this project is forewarning of machine failures in critical equipment at next-generation nuclear power plants (NPP). Test data were provided by two collaborating institutions: Duke Engineering and Services (first project year), and the Pennsylvania State University (Applied Research Laboratory) during the second and third project years. New nonlinear methods were developed and applied successfully to extract forewarning trends from process-indicative, time-serial data for timely, condition-based maintenance. Anticipation of failures in critical equipment at next-generation NPP will improve the scheduling of maintenance activities to minimize safety concerns, unscheduled non-productive downtime, and collateral damage due to unexpected failures. This approach provides significant economic benefit, and is expected to improve public acceptance of nuclear power. The approach is a multi-tiered, model-independent, and data-driven analysis that uses ORNL's novel nonlinear method to extract forewarning of machine failures from appropriate data. The first tier of the analysis provides a robust choice for the process-indicative data. The second tier rejects data of inadequate quality. The third tier removes signal artifacts that would otherwise confound the analysis, while retaining the relevant nonlinear dynamics. The fourth tier converts the artifact-filtered time-serial data into a geometric representation, that is then transformed to a discrete distribution function (DF). This method allows for noisy, finite-length datasets. The fifth tier obtains dissimilarity measures (DM) between the nominal-state DF and subsequent test-state DFs. Forewarning of a machine failure is indicated by several successive occurrences of the DM above a threshold, or by a statistically significant trend in the DM. This paradigm yields robust nonlinear signatures of degradation and its progression, allowing earlier and more accurate detection of the machine failure.

  6. The Role of Suomi NPP VIIRS Data in Land Science and Applications.

    Justice, C. O.; Csiszar, I. A.; Roman, M. O.; Vermote, E.

    2014-12-01

    The current Suomi-NPP mission was designed as a bridging mission to the Joint Polar Satellite System (JPSS) next-generation, operational satellites. The VIIRS instrument on-board Suomi-NPP provides continuity with NASA's Earth Observing System Moderate-resolution Imaging Spectroradiometer (MODIS) observations and a much-needed replacement of the long-serving, operational NOAA Advanced Very High Resolution Radiometer (AVHRR) system, to meet the expanding needs of Earth Science and Applications of Societal Benefit. Post-launch evaluation has proven the instrument to be excellent for land observations and capable of providing measurement continuity with MODIS. This provides a critical contribution to the scientific study of global change. Several regions of the World are undergoing rapid land transformations, driven by economic development and population growth. In addition, changes in climate conditions are resulting in ecosystem responses and changes in land cover and land use. Long-term, systematic observations of the global land surface at coarse resolution enable detection, monitoring and characterization of such changes to the land surface. A suite of Land environmental data records (EDRs) from the VIIRS, are being developed by NOAA to meet operational data needs, primarily for the National Weather Service (e.g., Albedo, Land Surface Temperature, Fractional Vegetation Cover, Surface Type, Snow and Ice Monitoring). Scientists funded by NOAA and NASA, have been evaluating and validating these products. Based on these evaluations, NASA is embarking on a program to develop enhanced and additional products to provide continuity with MODIS to meet the needs of the global change science community. In addition, and as with MODIS, data from the VIIRS can be used as input to a number of practical applications of societal benefit and the associated decision support systems. For example, progress with VIIRS data is being made in the areas of fire and agricultural monitoring

  7. Application of Mössbauer spectroscopy on corrosion products of NPP

    Dekan, J.; Lipka, J.; Slugeň, V.

    2013-04-01

    Steam generator (SG) is generally one of the most important components at all nuclear power plants (NPP) with close impact to safe and long-term operation. Material degradation and corrosion/erosion processes are serious risks for long-term reliable operation. Steam generators of four VVER-440 units at nuclear power plants V-1 and V-2 in Jaslovske Bohunice (Slovakia) were gradually changed by new original "Bohunice" design in period 1994-1998, in order to improve corrosion resistance of SGs. Corrosion processes before and after these design and material changes in Bohunice secondary circuit were studied using Mössbauer spectroscopy during last 25 years. Innovations in the feed water pipeline design as well as material composition improvements were evaluated positively. Mössbauer spectroscopy studies of phase composition of corrosion products were performed on real specimens scrapped from water pipelines or in form of filters deposits. Newest results in our long-term corrosion study confirm good operational experiences and suitable chemical regimes (reduction environment) which results mostly in creation of magnetite (on the level 70 % or higher) and small portions of hematite, goethite or hydrooxides. Regular observation of corrosion/erosion processes is essential for keeping NPP operation on high safety level. The output from performed material analyses influences the optimisation of operating chemical regimes and it can be used in optimisation of regimes at decontamination and passivation of pipelines or secondary circuit components. It can be concluded that a longer passivation time leads more to magnetite fraction in the corrosion products composition.

  8. NPP VIIRS On-Orbit Calibration and Characterization Using the Moon

    Sun, J.; Xiong, X.; Butler, J.

    2012-01-01

    The Visible Infrared Imager Radiometer Suite (VIIRS) is one of five instruments on-board the Suomi National Polar orbiting Partnership (NPP) satellite that launched from Vandenberg Air Force Base, Calif., on Oct. 28, 2011. VIIRS has been scheduled to view the Moon approximately monthly with a spacecraft roll maneuver after its NADIR door open on November 21, 2011. To reduce the uncertainty of the radiometric calibration due to the view geometry, the lunar phase angles of the scheduled lunar observations were confined in the range from -56 deg to -55 deg in the first three scheduled lunar observations and then changed to the range from -51.5 deg to -50.5 deg, where the negative sign for the phase angles indicates that the VIIRS views a waxing moon. Unlike the MODIS lunar observations, most scheduled VIIRS lunar views occur on the day side of the Earth. For the safety of the instrument, the roll angles of the scheduled VIIRS lunar observations are required to be within [-14 deg, 0 deg] and the aforementioned change of the phase angle range was aimed to further minimize the roll angle required for each lunar observation while keeping the number of months in which the moon can be viewed by the VIIRS instrument each year unchanged. The lunar observations can be used to identify if there is crosstalk in VIIRS bands and to track on-orbit changes in VIIRS Reflective Solar Bands (RSB) detector gains. In this paper, we report our results using the lunar observations to examine the on-orbit crosstalk effects among NPP VIIRS bands, to track the VIIRS RSB gain changes in first few months on-orbit, and to compare the gain changes derived from lunar and SD/SDSM calibration.

  9. The International Cooperation and Partnership, Keystones for Engineering and Procurement for Cernavoda NPP Unit 2

    Stiopol, Mihaela; Vatamanu, Mariana; Bucur, Cristina [Nuclearelectrica SA, 65 Polona Street, 010494 Bucharest (Romania)

    2008-07-01

    Romania is starting the second year as a MS of the EU, and is crossing a crucial year in implementing the national energy strategy. The EC documents from last year, as an energy policy for Europe or the nuclear illustrative programme provided the guidelines for the new Romanian energy strategy, based on the principles of security of supply, competitiveness, environment protection and optimization of the use of domestic natural resources. In the line with the right of each European state to choose the proper energy mix, in order to face the predicted deployment of natural gas and oil resources, Romania has to redirect its strategy to improve the energy efficiency, develop the renewable energies and to extend the use of nuclear. Romania has developed the national infrastructure for implementation, management and supervision of the nuclear power projects, including the environment protection aspects. The Ministry of Economy and Finance is responsible for the national strategy in the energy field, including NPP projects implementation and operation. Other important actors are the Ministry of the Environment, the Nuclear Agency, the Nuclear Regulatory Body - CNCAN and the National Agency for Radioactive Waste - ANDRAD. For Romania, the nuclear energy represents an obvious reality, strongly proved by good performances registered by Cernavoda NPP Unit 1 during over 12 years of commercial operation: High quality of the nuclear fuel produced by Nuclear Fuel Plant in Pitesti - as demonstrated by the 7 consecutive years of operation of Cernavoda Unit 1 without detecting a single fuel bundle with nuclear defect, the very good quality of heavy water produced in Romania based on Romanian concept and technology and, by the most recently success of the Romanian nuclear industry: commissioning of the second Unit at Cernavoda NPP on October 5. 2007. All those represent achievements of entire Romanian society for which the education, research, design engineering and operation of

  10. A process model for simulating net primary productivity (NPP) based on the interaction of water-heat process and nitrogen: a case study in Lantsang valley

    ZHANG Hai-long; LIU Gao-huan; FENG Xian-feng

    2011-01-01

    Terrestrial carbon cycle and the global atmospheric CO2 budget are important foci in global climate change research. Simulating net primary productivity (NPP) of terrestrial ecosystems is important for carbon cycle research. In this study, a plant-atmosphere-soil continuum nitrogen (N) cycling model was developed and incorporated into the Boreal Ecosystem Productivity Simulator (BEPS) model. With the established database (leaf area index, land cover, daily meteorology data,vegetation and soil) at a 1 km resolution, daily maps of NPP for Lantsang valley in 2007 were produced, and the spatial-temporal patterns of NPP and mechanisms of its responses to soil N level were further explored.The total NPP and mean NPP of Lantsang valley in 2007 were 66.5 Tg C and 416 g·m-2·a-1 C, respectively. In addition, statistical analysis of NPP of different land cover types was conducted and investigated. Compared with BEPS model (without considering nitrogen effect), it was inferred that the plant carbon fixing for the upstream of Lantsang valley was also limited by soil available nitrogen besides temperature and precipitation.However, nitrogen has no evident limitation to NPP accumulation of broadleaf forest, which mainly distributed in the downstream of Lantsang valley.

  11. Qualitative analysis of the maintenance politics of the systems of a typical PWR by artificial neural networks; Analise qualitativa da politica de manutencoes dos sistemas de um PWR tipico por redes neurais artificiais

    Lourenco, Victor Hugo Moreno

    2010-02-15

    Proceedings and techniques in order to maximize the reliability and the availability of industrial plants have been used along the last decades by specialists and professionals of maintenance. However, the modem industrial systems' sizing, and the increasing complexity and interdependence among its components have become this activity's planning a more and more difficult task. Considering this scenario, the objective of the present work is to provide a computational tool which is able to help about the taking decision's task, and about planning policies of maintenance practiced in thermonuclear plants. The tool developed is based on the artificial neural networks (ANN) for the recognition of standards and establishment of correlations among events occurred in the components of pressurized water reactor (PWR) typical systems. The ANN work as miners of database of failure events, and are able to identify connections and to establish imperceptible inferences even for the most experienced specialists in maintenance of nuclear systems. The results were attained from realistic data and are confronted against the maintenance's classic policies which are practiced nowadays on PWR thermonuclear plants. These results show the solidity of the technique in valuing and predicting failures in a real power plant, and is able to be used as a tool for supporting decisions about planning maintenance policies on a typical PWR. (author)

  12. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall.

  13. Calculation of sample problems related to two-phase flow blowdown transients in pressure relief piping of a PWR pressurizer

    Shin, Y.W.; Wiedermann, A.H.

    1984-02-01

    A method was published, based on the integral method of characteristics, by which the junction and boundary conditions needed in computation of a flow in a piping network can be accurately formulated. The method for the junction and boundary conditions formulation together with the two-step Lax-Wendroff scheme are used in a computer program; the program in turn, is used here in calculating sample problems related to the blowdown transient of a two-phase flow in the piping network downstream of a PWR pressurizer. Independent, nearly exact analytical solutions also are obtained for the sample problems. Comparison of the results obtained by the hybrid numerical technique with the analytical solutions showed generally good agreement. The good numerical accuracy shown by the results of our scheme suggest that the hybrid numerical technique is suitable for both benchmark and design calculations of PWR pressurizer blowdown transients.

  14. Application of RELAP5/MOD1 for calculation of safety and relief valve discharge piping hydrodynamic loads. Final report. [PWR

    1982-12-01

    A series of operability tests of spring-loaded safety valves was performed at Combustion Engineering in Windsor, CT as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of PWR Utilities in response to the recommendations of NUREG-0578 and the requirements of the NRC. Experimental data from five of the safety valve tests are compared with RELAP5/MOD1 calculations to evaluate the capability of the code to determine the fluid-induced transient loads on downstream piping. Comparisons between data and calculations are given for transients with discharge of steam, water, and water loop seal followed by steam. RELAP5/MOD1 provides useful engineering estimates of the fluid-induced piping loads for all cases.

  15. PFM Analysis for Pre-Existing Cracks on Alloy 182 Weld in PWR Primary Water Environment using Monte Carlo Simulation

    Park, Jae Phil; Bahn, Chi Bum [Pusan National University, Busan (Korea, Republic of)

    2015-10-15

    Probabilistic Fracture Mechanics (PFM) analysis was generally used to consider the scatter and uncertainty of parameters in complex phenomenon. Weld defects could be present in weld regions of Pressurized Water Reactors (PWRs), which cannot be considered by the typical fracture mechanics analysis. It is necessary to evaluate the effects of the pre-existing cracks in welds for the integrity of the welds. In this paper, PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out using a Monte Carlo simulation. PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out. It was shown that inspection decreases the gradient of the failure probability. And failure probability caused by the pre-existing cracks was stabilized after 15 years of operation time in this input condition.

  16. Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

    Oettingen Mikołaj

    2015-09-01

    Full Text Available In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238 and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242. The main results were presented as a calculated-to-experimental ratio (C/E for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55. The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

  17. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  18. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Curca-Tiving, F.; Opel, S.

    2014-07-01

    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  19. ANALISIS SENSITIVITAS TURBULENSI ALIRAN PADA KANAL BAHAN BAKAR PWR BERBASIS CFD

    Endiah Puji Hastuti

    2015-04-01

    Full Text Available Turbulensi aliran pendingin pada proses perpindahan panas berfungsi untuk meningkatkan nilai koefisien perpindahan panas, tidak terkecuali aliran dalam kanal bahan bakar. Program CFD (CFD=computational fluid dynamics, FLUENT adalah program komputasi berbasis elemen hingga (finite element yang mampu memprediksi dan menganalisis fenomena dinamika aliran fluida secara teliti. Program perhitungan CFD dipilih dalam penelitian ini karena selain akurat juga dapat memberikan visualisasi dengan baik. Penelitian ini bertujuan untuk memahami karakteristika perpindahan panas, massa dan momentum dari dinding rod bahan bakar ke pendingin secara visual, pada medan temperatur, medan tekanan, dan medan energi kinetika pendingin, sebagai fungsi dinamika aliran di dalam kanal, pada kondisi tunak dan transien. Analisis dinamika aliran pada kanal bahan bakar PWR berbasis CFD dilakukan dengan menggunakan sampel data reaktor PWR dengan daya 1000 MWe dengan susunan bahan bakar 17x17. Untuk menguji sensitivitas persamaan aliran yang sesuai dengan model aliran turbulen pada kanal bahan bakar dilakukan pemodelan dengan menggunakan persamaan k-omega (Ƙ-ω, k-epsilon (Ƙ-ε, dan Reynold stress model (RSM. Pada analisis sensitivitas aliran turbulen di dalam kanal digunakan model mesh hexahedral dengan memilih tiga geometri sel yang masing masing berukuran 0,5 mm; 0,2 mm dan 0,15 mm. Hasil analisis menunjukkan bahwa pada analisis kondisi tunak (steady state, terdapat hasil yang mirip pada model turbulen Ƙ-ε standard dan Ƙ-ω standard. Pengujian terhadap kriteria Dittus Boelter untuk bilangan Nusselt menunjukkan bahwa model Reynold stress model (RSM direkomendasikan. Analisis sensitivitas terhadap geometri mesh antara sel yang berukuran 0,5 mm, 0,2 mm dan 0,15 mm, menunjukkan bahwa geometri sel sebesar 0,5 mm telah mencukupi. Aliran turbulen berkembang penuh telah tercapai pada model LES dan DES, meskipun hanya dalam waktu singkat (3 s, model LES memerlukan waktu komputasi

  20. Organ-specific gene expression in maize: The P-wr allele. Final report, August 15, 1993--August 14, 1996

    Peterson, T.A.

    1997-06-01

    The ultimate aim of our work is to understand how a regulatory gene produces a specific pattern of gene expression during plant development. Our model is the P-wr gene of maize, which produces a distinctive pattern of pigmentation of maize floral organs. We are investigating this system using a combination of classical genetic and molecular approaches. Mechanisms of organ-specific gene expression are a subject of intense research interest, as it is the operation of these mechanisms during eukaryotic development which determine the characteristics of each organism Allele-specific expression has been characterized in only a few other plant genes. In maize, organ-specific pigmentation regulated by the R, B, and Pl genes is achieved by differential transcription of functionally conserved protein coding sequences. Our studies point to a strikingly different mechanism of organ-specific gene expression, involving post-transcriptional regulation of the regulatory P gene. The novel pigmentation pattern of the P-wr allele is associated with differences in the encoded protein. Furthermore, the P-wr gene itself is present as a unique tandemly amplified structure, which may affect its transcriptional regulation.

  1. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  2. Disposal of Steam Generators from Decommissioning of PWR Nuclear Power Plants

    Walberg, Mirko; Viermann, Joerg; Beverungen, Martin [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, 45127 Essen (Germany); Kemp, Lutz [Kernkraftwerk Stade GmbH and Co.oHG, Bassenflether Chaussee, 21683 Stade (Germany); Lindstroem, Anders [Studsvik Nuclear AB, SE-611 82 Nykoeping (Sweden)

    2008-07-01

    Amongst other materials remarkable amounts of radioactively contaminated or activated scrap are generated from the dismantling of Nuclear Power Plants. These scrap materials include contaminated pipework, fittings, pumps, the reactor pressure vessel and other large components, most of them are heat exchangers. Taking into account all commercial and technical aspects an external processing and subsequent recycling of the material might be an advantageous option for many of these components. The disposal of steam generators makes up an especially challenging task because of their measures, their weight and compared to other heat exchangers high radioactive inventory. Based on its experiences from many years of disposal of smaller components of NPP still in operation or under decommissioning GNS and Studsvik Nuclear developed a concept for disposal of steam generators, also involving experiences made in Sweden. The concept comprises transport preparations and necessary supporting documents, the complete logistics chain, steam generator treatment and the processing of arising residues and materials not suitable for recycling. The first components to be prepared, shipped and treated according to this concept were four steam generators from the decommissioning of the German NPP Stade which were removed from the plant and shipped to the processing facility during the third quarter of 2007. Although the plant had undergone a full system decontamination, due to the remaining contamination in a number of plugged tubes the steam generators had to be qualified as industrial packages, type 2 (IP-2 packages), and according to a special requirement of the German Federal Office for Radiation Protection a license for a shipment under special arrangement had to be applied for. The presentation gives an overview of the calculations and evidences required within the course of the IP-2 qualification, additional requirements of the competent authorities during the licensing procedure as

  3. Characterization of PWR vessel steel tearing under severe accident condition temperatures

    Matheron, Philippe, E-mail: philippe.matheron@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Chapuliot, Stephane, E-mail: stephane.chapuliot@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Nicolas, Laetitia, E-mail: laetitia.nicolas@cea.fr [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Laboratoire de Mecanique des Structures Industrielles Durables, UMR CNRS-EDF 2832, 1 avenue du General de Gaulle, F-92141 Clamart (France); Koundy, Vincent, E-mail: vincent.koundy@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France); Caroli, Cataldo, E-mail: cataldo.caroli@irsn.fr [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer We characterized French PWR vessel steel tearing resistance at high temperatures. Black-Right-Pointing-Pointer Tearing tests on Compact Tension (CT) specimens were carried out. Black-Right-Pointing-Pointer The variability of tearing properties with PWR vessels specifications was studied. Black-Right-Pointing-Pointer We propose a tearing criterion (energy parameter Gfr) at high temperatures. - Abstract: In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment. On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance. The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens. The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead

  4. Improving NPP availability using thermalhydraulic integral plant models. Assessment and application of turbine run back scenarios

    Reventos, F. [ANACNV, l' Hospitalet de l' Infant, Tarragona (Spain)]|[Technical University of Catalonia, UPC (Spain); Llopis, C.; Pretel, C. [Technical University of Catalonia, UPC (Spain); Posada, J.M.; Moreno, P. [Pablo Moreno S.A. (Spain)

    2001-07-01

    ANAV is the utility responsible of Asco and Vandellos Nuclear Power Plants, a two-unit and a single unit 1000 MW PWR plant, respectively. Both plants, Asco and Vandellos, are in normal operation since 1983 and 1987 and have undergone different important improvements like: steam generators and turbine substitution, power up-rating... Best estimate simulation by means of the thermal-hydraulic integral models of operating nuclear power plants are today impressively helpful for utilities in their purpose of improving availability and keeping safety level. ANAV is currently using Relap5/mod3.2 models of both plants for different purposes related to safety, operation, engineering and training. Turbine run-back system is designed to avoid reactor trips, and it does so in the existing plants, when the key parameters are correctly adjusted. The fine adjustment of such parameters was traditionally performed following the results of control simulators. Such simulators used a fully developed set of control equations and a quite simplified thermal-hydraulic feed-back. Boundary scenarios were considered in order to overcome the difficulties generated by simplification. (author)

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

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

    2010-10-15

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

  6. Thermodynamic analysis of SCW NPP cycles with thermo-chemical co-generation of hydrogen

    Naidin, N.; Mokry, S.; Monichan, R.; Chophla, K.; Pioro, I. [Faculty of Energy Systems and Nuclear Science, Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)], E-mail: Maria.Naidin@mycampus.uoit.ca, Sarah.Mokry@mycampus.uoit.ca, Romson.Monichan@uoit.ca, Karan.Chophla@mycampus.uoit.ca, Igor.Pioro@uoit.ca; Naterer, G.; Gabriel, K. [Faculty of Engineering and Applied Science, Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)], E-mail: Greg.Naterer@uoit.ca, Kamiel.Gabriel@uoit.ca

    2009-07-01

    Research activities are currently conducted worldwide to develop Generation IV nuclear reactor concepts with the objective of improving thermal efficiency and increasing economic competitiveness of Generation IV Nuclear Power Plants (NPPs) compared to modern thermal power plants. The Super-Critical Water-cooled Reactor (SCWR) concept is one of the six Generation IV options chosen for further investigation and development in several countries including Canada and Russia. Water-cooled reactors operating at subcritical pressures (10 - 16 MPa) have provided a significant amount of electricity production for the past 50 years. However, the thermal efficiency of the current NPPs is not very high (30 - 35%). As such, more competitive designs, with higher thermal efficiencies, which will be close to that of modern thermal power plants (45 - 50%), need to be developed and implemented. Super-Critical Water (SCW) NPPs will have much higher operating parameters compared to current NPPs (i.e., steam pressures of about 25 MPa and steam outlet temperatures up to 625{sup o}C). Furthermore, SCWRs operating at higher temperatures can facilitate an economical co-generation of hydrogen through thermochemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis. The two SCW NPP cycles proposed by this paper are based on direct, regenerative, no-reheat and single-reheat configurations. As such, the main parameters and performance in terms of thermal efficiency of the SCW NPP concepts mentioned above are being analyzed. The cycles are generally comprised of: an SCWR, a SC turbine, one deaerator, ten feedwater heaters, and pumps. The SC turbine of the no-reheat cycle consists of one High-Pressure (HP) cylinder and two Low-Pressure (LP) cylinders. Alternatively, the SC turbine for the single-reheat cycle is comprised of one High-Pressure (HP) cylinder, one Intermediate-Pressure (IP) cylinder and two Low-Pressure (LP) cylinders. Since the single

  7. IPEEE Review of other external events of the NPP Asco; Revision del IPEEE de sucesos externos de C.N. Asco

    Canadell, F.; Aleman, A.; Beltran, F.; Pifarre, D.; Hernandez, H.; Gasca, C.

    2011-07-01

    Within the process of maintaining and updating the risk analysis of the NPP Asco, results from the review of the vulnerability study of the plant against severe accidents caused by external success (Individual Plant Examination of External Events, IPEEE).

  8. Net primary productivity (NPP) and associated parameters for the U.S. outer continental shelf waters, 1998-2009 (NODC Accession 0071184)

    National Oceanic and Atmospheric Administration, Department of Commerce — This accession consists of monthly net primary productivity (NPP) estimates for 1998-2009 derived from the Vertically Generalized Production Model (VGPM) for the 26...

  9. Regression tree modeling of forest NPP using site conditions and climate variables across eastern USA

    Kwon, Y.

    2013-12-01

    As evidence of global warming continue to increase, being able to predict forest response to climate changes, such as expected rise of temperature and precipitation, will be vital for maintaining the sustainability and productivity of forests. To map forest species redistribution by climate change scenario has been successful, however, most species redistribution maps lack mechanistic understanding to explain why trees grow under the novel conditions of chaining climate. Distributional map is only capable of predicting under the equilibrium assumption that the communities would exist following a prolonged period under the new climate. In this context, forest NPP as a surrogate for growth rate, the most important facet that determines stand dynamics, can lead to valid prediction on the transition stage to new vegetation-climate equilibrium as it represents changes in structure of forest reflecting site conditions and climate factors. The objective of this study is to develop forest growth map using regression tree analysis by extracting large-scale non-linear structures from both field-based FIA and remotely sensed MODIS data set. The major issue addressed in this approach is non-linear spatial patterns of forest attributes. Forest inventory data showed complex spatial patterns that reflect environmental states and processes that originate at different spatial scales. At broad scales, non-linear spatial trends in forest attributes and mixture of continuous and discrete types of environmental variables make traditional statistical (multivariate regression) and geostatistical (kriging) models inefficient. It calls into question some traditional underlying assumptions of spatial trends that uncritically accepted in forest data. To solve the controversy surrounding the suitability of forest data, regression tree analysis are performed using Software See5 and Cubist. Four publicly available data sets were obtained: First, field-based Forest Inventory and Analysis (USDA

  10. S-NPP CrIS Full Resolution Sensor Data Record Processing and Evaluations

    Chen, Y.; Han, Y.; Wang, L.; Tremblay, D. A.; Jin, X.; Weng, F.

    2014-12-01

    The Cross-track Infrared Sounder (CrIS) on Suomi National Polar-orbiting Partnership Satellite (S-NPP) is a Fourier transform spectrometer. It provides a total of 1305 channels in the normal mode for sounding the atmosphere. CrIS can also be operated in the full spectral resolution (FSR) mode, in which the MWIR and SWIR band interferograms are recorded with the same maximum path difference as the LWIR band and with spectral resolution of 0.625 cm-1 for all three bands (total 2211 channels). NOAA will operate CrIS in FSR mode in December 2014 and the Joint Polar Satellite System (JPSS). Up to date, the FSR mode has been commanded three times in-orbit (02/23/2012, 03/12/2013, and 08/27/2013). Based on CrIS Algorithm Development Library (ADL), CrIS full resolution Processing System (CRPS) has developed to generate the FSR Sensor Data Record (SDR). This code can also be run for normal mode and truncation mode SDRs with recompiling. Different calibration approaches are implemented in the code in order to study the ringing effect observed in CrIS normal mode SDR and to support to select the best calibration algorithm for J1. We develop the CrIS FSR SDR Validation System to quantify the CrIS radiometric and spectral accuracy, since they are crucial for improving its data assimilation in the numerical weather prediction, and for retrieving atmospheric trace gases. In this study, CrIS full resolution SDRs are generated from CRPS using the data collected from FSR mode of S-NPP, and the radiometric and spectral accuracy are assessed by using the Community Radiative Transfer Model (CRTM) and European Centre for Medium-Range Weather Forecasts (ECMWF) forecast fields. The biases between observation and simulations are evaluated to estimate the FOV-2-FOV variability and bias under clear sky over ocean. Double difference method and Simultaneous Nadir Overpass (SNO) method are also used to assess the CrIS radiance consistency with well-validated IASI. Two basic frequency validation

  11. Evaluation of NPP VIIRS Vegetation Index EDR performance using MODIS and AVHRR data records

    vargas, M.; Shabanov, N.; Miura, T.

    2012-12-01

    Vegetation Index (VI) is one key parameter to specify the boundary condition in global climate models, weather forecasting models and numerous remote sensing applications for monitoring environmental state and its change. The VI Environmental Data Record (EDR), which includes the Top of Atmosphere Normalized Difference Vegetation Index (TOA NDVI) and the Top of Canopy Enhanced Vegetation Index (TOC EVI), is currently operationally generated from data delivered by the Visible Infrared Imaging radiometer Suite (VIIRS) instrument onboard the National Polar-orbiting Partnership (NPP) platform launched in October 2011. The VI EDR was implemented to provide continuity for 30+ years of historical VI records provided by MODIS and AVHRR sensors. This presentation reports on the results of the analysis performed by the JPSS VI group at NOAA-NESDIS-STAR on two major aspects of performance of the VI EDR in the early phase of the NPP mission: (1) assessment of accuracy of the VIIRS VI EDR product with respect to input data including Surface Reflectances, Cloud and Aerosol masks as function of vegetation (biome) types; (2) temporal and spatial consistency of VIIRS VI EDR with respect to heritage MODIS and AVHRR VI products. This analysis is based on data from VIIRS (daily TOA NDVI and TOC EVI, and daily surface reflectances), Terra MODIS (16 days composites of TOC EVI and TOC NDVI, and daily TOA radiances) and NOAA-18 AVHRR (7-days composites of TOA NDVI). MODIS 8-biome landcover mask was used to quantify variations in VI product performance as function of vegetation type. Best overall agreement is achieved between VIIRS and MODIS data (TOC EVI and TOC NDVI) in terms of minimum systematic discrepancy (minimum bias and STD) and highest correlation of spatial patterns (highest r^2). The agreement is highest for biomes with low vegetation cover, but degrades with increased foliage density. VIIRS cloud mask provides a fair screening of daily data over the globe. While performance of

  12. Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET

    Kliem, S.

    1998-10-01

    Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)

  13. Simulated dynamics of net primary productivity (NPP) for outdoor livestock feeding coefficients driven by climate change scenarios in México

    A. I. MONTERROSO RIVAS; J. D. GÓMEZ DÍAZ; M. L. TOLEDO MEDRANO; J. A. TINOCO RUEDA; C. CONDE ÁLVAREZ; C. GAY GARCÍA

    2011-01-01

    In this paper the concept of Net Primary Productivity (NPP) is used as a way to estimate the capacity of the ecosystem to produce dry matter which may be available for livestock to meet the forage requirements. The method allows the simulation of the possible impact on NPP and dry matter (DM), under climate change conditions observable for the country in a given time horizon. The concept was also used for current coefficients of rangeland and under current climate change scenarios, thus allow...

  14. NPP changes of Larix chinensis estimated by tree-ring data and its response to climate change in the northern and southern slopes of Mt. Taibai, central China

    Fang, O.; Xuemei, S.

    2015-12-01

    Larix chinensis is mainly distributed in timberline of Mt. Taibai in the Qinling Mountains, a critical geographic demarcation for climate and vegetation distribution in China. Combined with biomass equations and the annual diameter at breast height calculated from tree-ring widths and investigation data of sampling plots, annual biomass and net primary productivity (NPP) of L. chinensis in northern and southern slopes were estimated. Correlation and response analyses were used to illustrate the relationship between the climate and NPP. The results show that from 1949 to 2014, the biomass of L. chinensis in the pure forests increases from 54.03 to 94.43 t/ha in the northern slope and 28.32 to 55.80 t/ha in the southern slope. The NPP of L. chinensis in northern and southern slopes has varied concordantly over the past 65 years, with an average value of 0.62 and 0.42 t/(ha·a) respectively. The difference in NPPs between the northern and southern slope is decreasing for the slight decrease trend of NPP in northern slope. Temperature plays an important role in the growth of L. chinensis. Low temperature before the growing seasons (from pervious November to April) and warm conditions in the growing seasons (mainly from June to July) can increase the growth of L. chinensis. However, the relationships between NPP and temperature are different in the northern and southern slope. The NPP in southern slope is more positively correlated with the temperature in the growing seasons and there is no significant correlation relationship between the NPP and the temperature in previous winter (from pervious November to January), while the NPP in northern slope is more negatively correlated with the temperature before the growing seasons. These results will provide useful information for the future research of forest carbon cycling.

  15. Replacement of Co-base alloy for radiation exposure reduction in the primary system of PWR

    Han, Jeong Ho; Nyo, Kye Ho; Lee, Deok Hyun; Lim, Deok Jae; Ahn, Jin Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kim, Sun Jin [Hanyang Univ., Seoul (Korea, Republic of)

    1996-01-01

    Of numerous Co-free alloys developed to replace Co-base stellite used in valve hardfacing material, two iron-base alloys of Armacor M and Tristelle 5183 and one nickel-base alloy of Nucalloy 488 were selected as candidate Co-free alloys, and Stellite 6 was also selected as a standard hardfacing material. These four alloys were welded on 316SS substrate using TIG welding method. The first corrosion test loop of KAERI simulating the water chemistry and operation condition of the primary system of PWR was designed and fabricated. Corrosion behaviors of the above four kinds of alloys were evaluated using this test loop under the condition of 300 deg C, 1500 psi. Microstructures of weldment of these alloys were observed to identify both matrix and secondary phase in each weldment. Hardnesses of weld deposit layer including HAZ and substrate were measured using micro-Vickers hardness tester. The status on the technology of Co-base alloy replacement in valve components was reviewed with respect to the classification of valves to be replaced, the development of Co-free alloys, the application of Co-free alloys and its experiences in foreign NPPs, and the Co reduction program in domestic NPPs and industries. 18 tabs., 20 figs., 22 refs. (Author).

  16. Irradiation Effects Test Series: Test IE-2. Test results report. [PWR

    Allison, C. M.; Croucher, D. W.; Ploger, S. A.; Mehner, A. S.

    1977-08-01

    The report describes the results of a test using four 0.97-m long PWR-type fuel rods with differences in diametral gap and cladding irradiation. The objective of this test was to provide information about the effects of these differences on fuel rod behavior during quasi-equilibrium and film boiling operation. The fuel rods were subjected to a series of preconditioning power cycles of less than 30 kW/m. Rod powers were then increased to 68 kW/m at a coolant mass flux of 4900 kg/s-m/sup 2/. After one hour at 68 kW/m, a power-cooling-mismatch sequence was initiated by a flow reduction at constant power. At a flow of 2550 kg/s-m/sup 2/, the onset of film boiling occurred on one rod, Rod IE-011. An additional flow reduction to 2245 kg/s-m/sup 2/ caused the onset of film boiling on the remaining three rods. Data are presented on the behavior of fuel rods during quasiequilibrium and during film boiling operation. The effects of initial gap size, cladding irradiation, rod power cycling, a rapid power increase, and sustained film boiling are discussed. These discussions are based on measured test data, preliminary postirradiation examination results, and comparisons of results with FRAP-T3 computer model calculations.

  17. Performance of monosphere new gel type ion exchange resins for condensate polisher at PWR plants

    Nakanishi, S.; Nakamura, M.; Asou, K. [Kansai Electric Power Co., Inc., Osaka (Japan); Izumi, T.; Deguchi, T.; Ino, T.; Hagiwara, M.

    1998-12-31

    There are two kinds of ion exchange resins of gel type and porous one which are used as condensate polisher in LWR nuclear power plants. In order to estimate the performance of these resins on the condensate polisher at the secondary cycle of Japanese PWR plants, a column test was performed setting the column test device in Ohi power station unit 1 of the Kansai Electric Power Co., Inc. and the variations of the resin properties and the samples at the end of column were analyzed. The column test showed that the cross-linking degree of the new gel resins used was lower than those of porous ones. The new resins captured larger amounts of Matrix-Diffused Crud than the conventional cation resins before regeneration but not after that. Whereas the surface adsorbed crud was less captured by the new resins than conventional anion resins. However, there were little differences among these resins in respects of rinsing characteristics, sphericity, water quality, break through capacity, etc. At the condensate polisher in the secondary system it was confirmed that new gel resins had almost the same performance as one of the conventional ones and could be applied to the actual plant. (M.N.)

  18. Test requirements for the integral effect test to simulate Korean PWR plants

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  19. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  20. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  1. Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

    Vasiliev, A.; Ferroukhi, H.; Zhu, T.; Pautz, A.

    2014-04-01

    The development of a high-fidelity computational scheme to estimate the accumulated fluence at the tips of PWR control rods (CR) has been initiated at the Paul Scherrer Institut (PSI). Both the fluence from high-energy (E>1 MeV) neutrons as well as for the thermal range (EMCNPX model that is then used to compute the neutron fluxes. A particular aspect that needs scrutiny is the ability of the MCNPX-based calculation methodology to accurately predict the flux shapes along the control rod surfaces, especially for fully withdrawn CRs. In that case, the tip is located a short distance above the core/reflector interface and since this situation corresponds to a large part of reactor operation, the accumulated fluence will highly depend on the achieved calculation accuracy and precision in this non-fueled zone. The objective of the work presented in this paper is to quantify the influence of nuclear data on the calculated fluxes at the CR tips by (1) conducting a systematic comparison of modern neutron cross-section libraries, including JENDL-4.0, JEFF-3.1.1 and ENDF/B-VII.0, and (2) by quantifying the uncertainties in the neutron flux calculations with the help of available neutron cross-section variances/covariances data. For completeness, the magnitude of these nuclear data-based uncertainties is also assessed in relation to the influence from other typical sources of modeling uncertainties/biases.

  2. Fatigue-crack growth behavior of Type 347 stainless steels under simulated PWR water conditions

    Hong, Seokmin; Min, Ki-Deuk; Yoon, Ji-Hyun; Kim, Min-Chul; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Fatigue crack growth rate (FCGR) curve of stainless steel exists in ASME code section XI, but it is still not considering the environmental effects. The longer time nuclear power plant is operated, the more the environmental degradation issues of materials pop up. There are some researches on fatigue crack growth rate of S304 and S316, but researches of FCGR of S347 used in Korea nuclear power plant are insufficient. In this study, the FCGR of S347 stainless steel was evaluated in the PWR high temperature water conditions. The FCGRs of S347 stainless steel under pressurized-water conditions were measured by using compact-tension (CT) specimens at different levels of dissolved oxygen (DO) and frequency. 1. FCGRs of SS347 were slower than that in ASME XI and environmental effect did not occur when frequency was higher than 1Hz. 2. Fatigue crack growth is accelerated by corrosion fatigue and it is more severe when frequency is slower than 0.1Hz. 3. Increase of crack tip opening time increased corrosion fatigue and it deteriorated environmental fatigue properties.

  3. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  4. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  5. Validation of the scale system for PWR spent fuel isotopic composition analyses

    Hermann, O.W.; Bowman, S.M.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Laboratories, Las Vegas, NV (United States)

    1995-03-01

    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant.

  6. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)

    2001-03-15

    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

  7. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  8. Evaluation of Fuel Performance Uncertainty in a PWR HFP RIA Analysis

    Lee, Joosuk; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    Sensitivity and combined uncertainty studies based on the various kinds of uncertainty sources have been carried out in a PWR hot full power (HFP) condition. - Cladding inner diameter, fuel thermal conductivity, fuel thermal expansion and peak power have induced a significant impact to the fuel enthalpy and temperature. - Cladding hoop strain was strongly affected by the uncertainty parameters of cladding inner diameter, fuel thermal expansion, EPRI-1 CHF and peak power. - Above results are valid in the given analysis condition in this paper. Thereby, the analysis conditions, for example the peak linear heat rate before RIA or peak power and FWHM etc, are changed the results will be changed also. Approved analysis methodology for licensing application in the safety analysis of reactivity initiated accident (RIA) in Korea is based on a conservative approach. But newly introduced safety criteria, described in section 4.2 of NUREG-0800, tend to reduce the margins or depending on the reactor types rod failure is predicted due to the pellet-to-cladding mechanical interaction (PCMI) criteria. Thereby, licensee is trying to improve the margins by utilizing a less conservative approach.

  9. PWR composite materials use. A particular case of safety-related service water pipes

    Pays, M.F.; Le Courtois, T

    1997-11-01

    This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during `lifetime`); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author) 2 refs.

  10. VOF Calculations of Countercurrent Gas-Liquid Flow in a PWR Hot Leg

    M. Murase

    2012-01-01

    Full Text Available We improved the computational grid and schemes in the VOF (volume of fluid method with the standard − turbulent model in our previous study to evaluate CCFL (countercurrent flow limitation characteristics in a full-scale PWR hot leg (750 mm diameter, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa. In this paper, therefore, to evaluate applicability of the VOF method to different fluid properties and a different scale, we did numerical simulations for full-scale air-water conditions and the 1/15-scale air-water tests (50 mm diameter, respectively. The results calculated for full-scale conditions agreed well with CCFL data and showed that CCFL characteristics in the Wallis diagram were mitigated under 1.5 MPa steam-water conditions comparing with air-water flows. However, the results calculated for the 1/15-scale air-water tests greatly underestimated the falling water flow rates in calculations with the standard − turbulent model, but agreed well with the CCFL data in calculations with a laminar flow model. This indicated that suitable calculation models and conditions should be selected to get good agreement with data for each scale.

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

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

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

  12. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  13. Analysis of measured and calculated counterpart test data in PWR and VVER 1000 simulators

    d’Auria Francesco

    2005-01-01

    Full Text Available This paper presents an over view of the "scaling strategy", in particular the role played by the counter part test methodology. The recent studies dealing with a scaling analysis in light water reactor with special regard to the VVER 1000 Russian reactor type are presented to demonstrate the phenomena important for scaling. The adopted scaling approach is based on the selection of a few characteristic parameters chosen by taking into account their relevance in the behavior of the transient. The adopted computer code used is RELAP5/Mod3.3 and its accuracy has been demonstrated by qualitative and quantitative evaluation. Comparing experimental data, it was found that the investigated facilities showed similar behavior concerning the time trends, and that the same thermal hydraulic phenomena on a qualitative level could be predicted. The main results are: PSB and LOBI main parameters have similar trends. This fact is the confirmation of the validity of the adopted scaling approach and it shows that PWR and VVER reactor type behavior is very similar. No new phenomena occurred during the counter part test, despite the fact that the two facilities had a different lay out, and the already known phenomena were predicted correctly by the code. The code capability and accuracy are scale-independent. Both character is tics are necessary to permit the full scale calculation with the aim of nuclear power plant behavior prediction. .

  14. Fatigue Crack Growth Rate Behavior of Type 347 Stainless Steel in Simulated PWR Water Environment

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The pressurizer surge line of a Korean standard nuclear power plane uses Nb stabilized type 347 stainless steel. The pressurizer surge line is the pipe connecting the pressurizer and the hot leg line, and the path controlling the pressure and temperature of the cooling system of the nuclear reactor, operated at 316 .deg. C and in a 150atm. The pressurizer surge line operated at high temperature and high pressure receives thermal stress by a temperature change and mechanical stress by a pressure change at the same time, and by being exposed to the high temperature and high pressure cooling water environment of a nuclear power plant, environmental fatigue by stress and corrosion is the main damage instrument. As the effect of environmental fatigue has been reported, through low cycle fatigue, fatigue life evaluations of austenite stainless steel have been conducted, but evaluations of fatigue crack growth rate to evaluate the soundness are very poor. In this study, evaluated characteristics of fatigue crack growth rate base on a change of dissolved oxygen in a PWR environment

  15. Development of a parametric containment event tree model of a severe PWR accident

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

    1996-06-01

    The study supports the development project of STUK on `Living` PSA Level 2. The main work objective is to develop review tools for the Level 2 PSA studies underway at the utilities. The SPSA (STUK PSA) code is specifically designed for the purpose. In this work, SPSA is utilized as the Level 2 programming and calculation tool. A containment event tree (CET) model is built for analysis of severe accidents at the Loviisa pressurized water reactor (PWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to include new research results, and so it facilitates the Living PSA concept on Level 2 as well. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting at a low primary system pressure. Severe accident progression from five plant damage states (PDSs) is examined, however the integration with Level 1 is deferred to more definitive, integrated, safety assessments. (34 refs., 5 figs., 9 tabs.).

  16. MELCOR 1.8.2 assessment: Surry PWR TMLB` (with a DCH study)

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

    1994-02-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified.

  17. RADIATION SITUATION STATUS OF THE TERRITORY OF OREL REGION, AFFECTED BY THE RADIOACTIVE CONTAMINATION DUE TO CHERNOBYL NPP ACCIDENT

    G. L. Zakharchenko

    2009-01-01

    Full Text Available The article contains analysis of activities of Centers of State Sanitary and Epidemiological Inspection in Orel region during the first days after the Chernobyl NPP accident and consecutive years. The results of multi-year radiation and hygienic monitoring on the territory of Orel region are presented; efficiency of various measures of exposure dose reduction for the population of Orel region is analyzed.

  18. TRANSBOUNDARY IMPACT OF THE CHERNAVODSKA NPP ON TRITIUM POLLUTION OF THE DANUBE RIVER ON THE TERRITORY OF UKRAINE

    V. I. VIT`KO

    2015-10-01

    Full Text Available This article reviews the influence of the Chernavodska nuclear power plant on the aquatic environment of the Danube River in the transboundary context. Data of tritium discharges, dependence of volume activity of tritium in the Danube River, and its inflows from the surrounding areas to its mouth. The average annual volume activities of tritium are provided. Assessments of the impact of the Chernavodska NPP in conditions that are different from the norm have been given.

  19. The new revision of NPP Krsko decommissioning, radioactive waste and spent fuel management program: analyses and results

    Zeleznik, Nadja; Kralj, Metka [ARAO, Parmova 53, 1000 Ljubljana (Slovenia); Lokner, Vladimir; Levanat, Ivica; Rapic, Andrea [APO, Savska 41, Zagreb (Croatia); Mele, Irena [IAEA, Vienna (Austria)

    2010-07-01

    The preparation of the new revision of the Decommissioning and Spent Fuel (SF) and Low and Intermediate level Waste (LILW) Disposal Program for the NPP Krsko (Program) started in September 2008 after the acceptance of the Term of Reference for the work by Intergovernmental Committee responsible for implementation of the Agreement between the governments of Slovenia and Croatia on the status and other legal issues related to investment, exploitation, and decommissioning of the Nuclear power plant Krsko. The responsible organizations, APO and ARAO together with NEK prepared all new technical and financial data and relevant inputs for the new revision in which several scenarios based on the accepted boundary conditions were investigated. The strategy of immediate dismantling was analyzed for planned and extended NPP life time together with linked radioactive waste and spent fuel management to calculate yearly annuity to be paid by the owners into the decommissioning funds in Slovenia and Croatia. The new Program incorporated among others new data on the LILW repository including the costs for siting, construction and operation of silos at the location Vrbina in Krsko municipality, the site specific Preliminary Decommissioning Plan for NPP Krsko which included besides dismantling and decontamination approaches also site specific activated and contaminated radioactive waste, and results from the referenced scenario for spent fuel disposal but at very early stage. Important inputs for calculations presented also new amounts of compensations to the local communities for different nuclear facilities which were taken from the supplemented Slovenian regulation and updated fiscal parameters (inflation, interest, discount factors) used in the financial model based on the current development in economical environment. From the obtained data the nominal and discounted costs for the whole nuclear program related to NPP Krsko which is jointly owned by Slovenia and Croatia have

  20. Accessing the impacts of bamboo expansion on NPP and N cycling in evergreen broadleaved forest in subtropical China

    Song, Qing-Ni; Lu, Hui; Liu, Jun; Yang, Jun; Yang, Guang-Yao; Yang, Qing-Pei

    2017-01-01

    Bamboo (Phyllostachys pubescens) expansion into adjacent forests is a widespread phenomenon in subtropical regions, and it has greatly changed the dominance hierarchy from trees to bamboos. This process may be accompanied by changes in productivity, nutrients accumulation and biogeochemical cycles. We compared the net primary production (NPP) and major pools and fluxes of nitrogen (N) in bamboo-dominant forest (BDF) and neighboring secondary evergreen broadleaved forest (EBF) in South China using the space-for-time substitution method. We found that the mean NPP of the BDF was 30.0 t ha‑1 yr‑1, which was 51.5% greater than that of the EBF (19.8 t ha‑1 yr‑1). The plant N pool for the BDF was 37.5% larger than that of the EBF, whereas the soil inorganic N pool significantly decreased by 31.2% with conversion of the EBF to BDF. Additionally, the ratio of N return to N uptake was 0.69 in the BDF and 0.88 in the EBF because of the lower litter N return of the BDF compared with that of the EBF. These results indicated that the expansion of P. pubescens significantly increased the NPP and plant N accumulation but reduced the soil N available pool and slowed the N cycling rate, which could lead to soil degradation.

  1. Accessing the impacts of bamboo expansion on NPP and N cycling in evergreen broadleaved forest in subtropical China

    Song, Qing-ni; Lu, Hui; Liu, Jun; Yang, Jun; Yang, Guang-yao; Yang, Qing-pei

    2017-01-01

    Bamboo (Phyllostachys pubescens) expansion into adjacent forests is a widespread phenomenon in subtropical regions, and it has greatly changed the dominance hierarchy from trees to bamboos. This process may be accompanied by changes in productivity, nutrients accumulation and biogeochemical cycles. We compared the net primary production (NPP) and major pools and fluxes of nitrogen (N) in bamboo-dominant forest (BDF) and neighboring secondary evergreen broadleaved forest (EBF) in South China using the space-for-time substitution method. We found that the mean NPP of the BDF was 30.0 t ha−1 yr−1, which was 51.5% greater than that of the EBF (19.8 t ha−1 yr−1). The plant N pool for the BDF was 37.5% larger than that of the EBF, whereas the soil inorganic N pool significantly decreased by 31.2% with conversion of the EBF to BDF. Additionally, the ratio of N return to N uptake was 0.69 in the BDF and 0.88 in the EBF because of the lower litter N return of the BDF compared with that of the EBF. These results indicated that the expansion of P. pubescens significantly increased the NPP and plant N accumulation but reduced the soil N available pool and slowed the N cycling rate, which could lead to soil degradation. PMID:28067336

  2. The development of robotic system for inspecting and repairing NPP primary coolant system of high-level radioactive environment

    Kim, Seung Ho; Kim, Ki Ho; Jung, Seung Ho; Kim, Byung Soo; Hwang, Suk Yeoung; Kim, Chang Hoi; Seo, Yong Chil; Lee, Young Kwang; Lee, Yong Bum; Cho, Jai Wan; Lee, Jae Kyung; Lee, Yong Deok

    1997-07-01

    This project aims at developing a robotic system to automatically handle inspection and maintenance of NPP safety-related facilities in high-level radioactive environment. This robotic system under development comprises two robots depending on application fields - a mobile robot and multi-functional robot. The mobile robot is designed to be used in the area of primary coolant system during the operation of NPP. This robot enables to overcome obstacles and perform specified tasks in unstructured environment. The multi-functional robot is designed for performing inspection and maintenance tasks of steam generator and nuclear reactor vessel during the overhaul periods of NPP. Nuclear facilities can be inspected and repaired all the time by use of both the mobile robot and the multi-functional robot. Human operator, by teleoperation, monitors the movements of such robots located at remote task environment via video cameras and controls those remotely generating desired commands via master manipulator. We summarize the technology relating to the application of the mobile robot to primary coolant system environment, the applicability of the mobile robot through 3D graphic simulation, the design of the mobile robot, the design of its radiation-hardened controller. We also describe the mechanical design, modeling, and control system of the multi-functional robot. Finally, we present the design of the force-reflecting master and the modeling of virtual task environment for a training simulator. (author). 47 refs., 16 tabs., 43 figs.

  3. A study on the Predictive Maintenance in NPP by Thermography diagnosis test

    Ryu, Gwangyeol; Lee, Goungjin [Chosun Univ., Kwangju (Korea, Republic of)

    2013-05-15

    The data are obtained, analyzed, trended, and used to predict equipment failures. When equipment failure timing is known, then actions can be taken to prevent or delay failure. This allows equipment reliability to remain high. Thermal measurement technology measures the absolute or relative temperatures of key equipment parts or areas being monitored. Abnormal temperatures indicate developing problems. Temperature and thermal behaviour of plant components are the most critical factors in the maintenance of plant equipment. For this reason, temperature is frequently considered the key to successful plant maintenance and is the frequently measured quantity. There are two types of equipment used in this technology: contact and non-contact. Contact methods of temperature measurement, using thermometers and thermocouples, are still commonly used for many applications. However, non-contact measurement using infrared sensors has become an increasingly desirable alternative to conventional methods. Maintenance is conducted every six months to two years on all the 428(in Young Gwang NPP 1, 2) plant thermography diagnosis equipment to be inspected including pumps, fans, transformers. Through the thermography measurement, there was a abnormal temperature detected recently on the transformers and pumps which was promptly responded to before showing any signs of malfunction, contributing to improving reliability of the facility. In case of the high-voltage switchgear, for the safety of the thermography inspector, additional safety precautions should be taken by installing the thermography measurement inspection window on the transformer's panel door.

  4. Experience in adjusting of the level regulation system of steam generators of the Rovno NPP

    Patselyuk, S.N.; Sokolov, A.G.; Kazakov, V.I.; Dorosh, Yu.A.

    1984-07-01

    A system of feed water level control in steam generators at the Rovno NPP with WWER-440 reactors which comprises start-up as well as main regulators is described. The start-up regulator (single-pulsed with a signal by the level) keeps the level in the steam generator at loadings up to 30% of the nominal reactor power Nsub(nom.). The main regulator is connected in the three-pulsed circuit and it receives signals by steam and water flow rate and by the level in the steam generator. The main regulator has been started only at loadings above 40% Nsub(nom.). After reconstruction it was used in the 15-100% Nsub(nom.) range. Characteristics of the level control system in the steam generator at perturbations intoduced by the main circulating pump (MCP) and turbine disconnection as well as change in feed water flow rate have been studied. The studies have revealed that the system ensures necessary quality of control in stationary modes. The system operates stably at perturbations of feed water flow rate up to 50% Nsub(nom.). Perturbations by MCP connections and disconnections is most difficult for control system.

  5. Analysis of radwaste management alternatives during dismantling of Ignalina NPP systems with low level contamination

    Poskas, Gintautas [Lithuanian Energy Institute, Kaunas (Lithuania). Nuclear Engineering Lab.; Kaunas Univ. of Technology (Lithuania); Poskas, Povilas; Simonis, Audrius [Lithuanian Energy Institute, Kaunas (Lithuania). Nuclear Engineering Lab.

    2013-12-15

    Ignalina NPP was operating two RBMK-1500 reactors which are under decommissioning now. In this paper, analysis on radwaste management alternatives during the dismantling of systems with low level contamination and different types of components in buildings 117/1 and V1 are presented. After situation analysis and collection of the primary information related to components' physical and radiological characteristics, location and other data, two alternatives for radwaste management during the dismantling were formulated and evaluated: the first one (A1) when the decontamination of the dismantled components is performed (if it is reasonable), and the second one (A2) when no decontamination of the dismantled components is performed and after the dismantling, the components are routed to appropriate waste storage or disposal sites. To select the preferable alternative, MCDA method - AHP (Analytic Hierarchy Process) is applied. Hierarchical lists of decision criteria, necessary for assessment of alternatives performance, are formulated. Quantitative decision criteria values for these alternatives are calculated using software DECRAD, which was developed by Lithuanian Energy Institute Nuclear Engineering Laboratory. Qualitative decision criteria are evaluated using expert judgment. Analysis results show that alternative A1 has a preference against alternative A2. (orig.)

  6. Westinghouse Fuel Assemblies Performance after Operation in South-Ukraine NPP Mixed Core

    Abdullayev, A. M.; Kulish, G. V.; Slyeptsov, O.; Slyeptsov, S.; Aleshin, Y.; Sparrow, S.; Lashevych, P.; Sokolov, D.; Latorre, Richard

    2013-09-14

    The evaluation of WWER-1000 Westinghouse fuel performance was done using the results of post–irradiation examinations of six LTAs and the WFA reload batches that have operated normally in mixed cores at South-Ukraine NPP, Unit-3 and Unit-2. The data on WFA/LTA elongation, FR growth and bow, WFA bow and twist, RCCA drag force and drag work, RCCA drop time, FR cladding integrity as well as the visual observation of fuel assemblies obtained during the 2006-2012 outages was utilized. The analysis of the measured data showed that assembly growth, FR bow, irradiation growth, and Zr-1%Nb grid and ZIRLO cladding corrosion lies within the design limits. The RCCA drop time measured for the LTA/WFA is about 1.9 s at BOC and practically does not change at EOC. The measured WFA bow and twist, and data of drag work on RCCA insertion showed that the WFA deformation in the mixed core is mostly controlled by the distortion of Russian FAs (TVSA) having the higher lateral stiffness. The visual inspection of WFAs carried out during the 2012 outages revealed some damage to the Zr-1%Nb grid outer strap for some WFAs during the loading sequence. The performed fundamental investigations allowed identifying the root cause of grid outer strap deformation and proposing the WFA design modifications for preventing damage to SG at a 225 kg handling trip limit.

  7. MORE. Management of Requirements in NPP modernisation projects - Project report 2005

    Thunem, A.P.J.; Fredriksen, R.; Thunem, H.P.J. [IFE (Norway); Ventae, O.; Valkonen, J.; Holmberg, J.E. [VTT Technical Research Centre of Finland (Finland)

    2006-04-15

    The overall objective of the project MORE is to improve the means for managing the large amounts of evolving requirements in Nordic NPP modernisation projects. In accordance to this objective, the activity will facilitate the industrial utilisation of the research results from the project TACO. On the basis of experiences in the Nordic countries, the overall aim of the TACO project has been to identify the best practices and most important criteria for ensuring effective communication in relation to requirements elicitation and analysis, understandability of requirements to all parties, and traceability of requirements. The project resulted in the development of a traceability model for handling requirements from their origins and through their final shapes. Particular emphasis for the MORE project in 2005 was put on utilising a prototype of a tool (TRACE) intended to support an adopted approach to dependable requirements engineering, suitable for modelling and handling large amounts of requirements related to all stages of the systems development process and not only those traditionally including requirements at high-level stages. (au)

  8. Development of Highly Survivable Power and Communication System for NPP Instruments under Severe Accident

    Yoo, Seung J.; Gu, Beom W.; Nguyen, Duy T.; Choi, Bo H.; Rim, Chun T. [KAIST, Daejeon (Korea, Republic of); Lee, So I. [KHNP CRI, Daejeon (Korea, Republic of)

    2014-10-15

    According to the detail report from the Fukushima nuclear accident, the failure of conventional instruments is mainly due to the following reasons. 1) Insufficient backup battery capacity after the station black out (SBO) 2) The malfunction or damage of instruments due to the extremely harsh ambient condition after the severe accident 3) The cut-off of power and communication cable due to the physical shocks of hydrogen explosion after the severe accident Since the current equipment qualification (EQ) for the NPP instruments is based on the design basis accident such as loss of coolant accident (LOCA), conventional instruments, which are examined under EQ condition, cannot guarantee their normal operation during the severe accident. A 7m-long-distance wireless power transfer and a radio frequency (RF) communication were introduced with conventional wired system to increase a redundancy. A heat isolation box and a harness are adopted to provide a protection from the expected physical shocks such as missiles and drastic increase of ambient temperature and pressure. A detail design principle of the highly survivable power and communication system, which has 4 sub-systems of a DCRS wireless power transfer, a Zigbee wireless communication, a GFRP harness, and a passive type router with a fly back regulator, has been presented in this paper. Each sub-system has been designed to have a robust operation characteristic regardless of the estimated physical shocks after the severe accident.

  9. Integrity Analysis of Turbine Building for the MSLB Using GOTHIC code for Wolsong NPP Unit 2

    Ko, Bong-Jin; Jin, Dong-Sik; Kim, Jong-Hyun; Han, Sang-Koo [ACT, Daejeon (Korea, Republic of); Choi, Hoon; Kho, Dong-Wook [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-05-15

    A break in the piping between the steam generators and the turbine can lead to rapid loss of secondary circuit inventory. A break inside the turbine building leads to pressure differentials between different areas of the turbine building. In order to improve the environmental protection of various components within the turbine building, a wall has been erected which effectively separates the area in which these components are housed from the rest of the turbine building. Relief panels installed in the turbine building ensure that the pressure differential across the wall would be less than that required to jeopardize the wall integrity. The turbine building service wing is excluded from the scope of this analysis. It is further assumed that any doors in the heavy wall are as strong as the wall itself, with no gaps or leakage around the doors. For the full scope safety analysis of turbine building for Wolsong NPP unit 2, input decks for the various objectives, which can be read by GOTHIC 7.2a, are developed and tested for the steady state simulation. The input data files provide simplified representations of the geometric layout of the turbine building (volumes, dimensions, flow paths, doors, panels, etc.) and the performance characteristics of the various turbine building subsystems.

  10. A Conceptual Framework of Human Reliability Analysis for Execution Human Error in NPP Advanced MCRs

    Jang, In Seok; Kim, Ar Ryum; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of); Jung, Won Dea [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-08-15

    The operation environment of Main Control Rooms (MCRs) in Nuclear Power Plants (NPPs) has changed with the adoption of new human-system interfaces that are based on computer-based technologies. The MCRs that include these digital and computer technologies, such as large display panels, computerized procedures, and soft controls, are called Advanced MCRs. Among the many features of Advanced MCRs, soft controls are a particularly important feature because the operation action in NPP Advanced MCRs is performed by soft control. Using soft controls such as mouse control, and touch screens, operators can select a specific screen, then choose the controller, and finally manipulate the given devices. Due to the different interfaces between soft control and hardwired conventional type control, different human error probabilities and a new Human Reliability Analysis (HRA) framework should be considered in the HRA for advanced MCRs. In other words, new human error modes should be considered for interface management tasks such as navigation tasks, and icon (device) selection tasks in monitors and a new framework of HRA method taking these newly generated human error modes into account should be considered. In this paper, a conceptual framework for a HRA method for the evaluation of soft control execution human error in advanced MCRs is suggested by analyzing soft control tasks.

  11. JACOS: AI-based simulation system for man-machine system behavior in NPP

    Yoshida, Kazuo; Yokobayashi, Masao; Tanabe, Fumiya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kawase, Katsumi [CSK Corp., Tokyo (Japan); Komiya, Akitoshi [Computer Associated Laboratory, Inc., Hitachinaka, Ibaraki (Japan)

    2001-08-01

    A prototype of a computer simulation system named JACOS (JAERI COgnitive Simulation system) has been developed at JAERI (Japan Atomic Energy Research Institute) to simulate the man-machine system behavior in which both the cognitive behavior of a human operator and the plant behavior affect each other. The objectives of this system development is to provide man-machine system analysts with detailed information on the cognitive process of an operator and the plant behavior affected by operator's actions in accidental situations of a nuclear power plant. The simulation system consists of an operator model and a plant model which are coupled dynamically. The operator model simulates an operator's cognitive behavior in accidental situations based on the decision ladder model of Rasmussen, and is implemented using the AI-techniques of the distributed cooperative inference method with the so-called blackboard architecture. Rule-based behavior is simulated using knowledge representation with If-Then type of rules. Knowledge-based behavior is simulated using knowledge representation with MFM (Multilevel Flow Modeling) and qualitative reasoning method. Cognitive characteristics of attentional narrowing, limitation of short-term memory, and knowledge recalling from long-term memory are also taken into account. The plant model of a 3-loop PWR is also developed using a best estimate thermal-hydraulic analysis code RELAP5/MOD2. This report is prepared as User's Manual for JACOS. The first chapter of this report describes both operator and plant models in detail. The second chapter includes instructive descriptions for program installation, building of a knowledge base for operator model, execution of simulation and analysis of simulation results. The examples of simulation with JACOS are shown in the third chapter. (author)

  12. Cloud-based Web Services for Near-Real-Time Web access to NPP Satellite Imagery and other Data

    Evans, J. D.; Valente, E. G.

    2010-12-01

    We are building a scalable, cloud computing-based infrastructure for Web access to near-real-time data products synthesized from the U.S. National Polar-Orbiting Environmental Satellite System (NPOESS) Preparatory Project (NPP) and other geospatial and meteorological data. Given recent and ongoing changes in the the NPP and NPOESS programs (now Joint Polar Satellite System), the need for timely delivery of NPP data is urgent. We propose an alternative to a traditional, centralized ground segment, using distributed Direct Broadcast facilities linked to industry-standard Web services by a streamlined processing chain running in a scalable cloud computing environment. Our processing chain, currently implemented on Amazon.com's Elastic Compute Cloud (EC2), retrieves raw data from NASA's Moderate Resolution Imaging Spectroradiometer (MODIS) and synthesizes data products such as Sea-Surface Temperature, Vegetation Indices, etc. The cloud computing approach lets us grow and shrink computing resources to meet large and rapid fluctuations (twice daily) in both end-user demand and data availability from polar-orbiting sensors. Early prototypes have delivered various data products to end-users with latencies between 6 and 32 minutes. We have begun to replicate machine instances in the cloud, so as to reduce latency and maintain near-real time data access regardless of increased data input rates or user demand -- all at quite moderate monthly costs. Our service-based approach (in which users invoke software processes on a Web-accessible server) facilitates access into datasets of arbitrary size and resolution, and allows users to request and receive tailored and composite (e.g., false-color multiband) products on demand. To facilitate broad impact and adoption of our technology, we have emphasized open, industry-standard software interfaces and open source software. Through our work, we envision the widespread establishment of similar, derived, or interoperable systems for

  13. Analysis of the containment of a compact reactor PWR submitted to loss of coolant accident; Analise da contencao de um reator PWR compacto submetido a acidente de perda de refrigerante

    Dutra, Alexandre de Souza; Belchior Junior, Antonio; Guimaraes, Leonam dos Santos [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2000-07-01

    In the present paper analyses were done with the computer code RELAP5/MOD2 for rising the process conditions of the containment of a compact reactor PWR of low potency, submitted to Loss of Coolant Accidents (LOCA). The main results obtained were the behavior of maximum conditions of pressure as a function of the available containment free volume. It was also studied the problem of containment sub-compartmentation, that is to say, the possibility of the rupture to happen in restricted spaces generating high sub-compartment peak pressure and, consequently, high strains on the internal structures. (author)

  14. Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR; Optimizacion de la distribucion de barras con oxido de gadolinio en elementos combustibles para reactores PWR

    Melgar Santa Cecilia, P. A.; Velazquez, J.; Ahnert Iglesias, C.

    2014-07-01

    In the schemes of low leakage, currently used in the majority of PWR reactors, it makes use of absorbent consumables for the effective control of the factors of peak, the critical concentration of initial boron and the moderator temperature coefficient. One of the most used absorbing is the oxide of gadolinium, which is integrated within the fuel pickup. Occurs a process of optimization of fuel elements with oxide of gadolinium, which allows for a smaller number of configurations with a low peak factor for bar. (Author)

  15. The safety analysis and thermohydraulic methodologies for the power updating analyses in Spanish PWR plants; Methodologias de diseno termohidraulico y de analisis de seguridad en los aumentos de potencia de centrales PWR

    Salesa, F.

    2014-02-01

    This article describes the Safety Analysis and Thermohydraulic methodologies used by ENUSA for the Power Updating analyses in Spanish PWR plants of Westinghouse design: Design tools have been developed over the first cycles resulting new correlations of DNB, fitted to the new fuel assemblies, new DNBR calculation methodology and other improvements in the design areas. Using these methodologies, the available margins between design and limit values are wider. These new margins have allowed to accomplish the design criteria under the new power updating operational conditions. (Author)

  16. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-08-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  17. Study for highly functional resin (macroporous resin) superior in removing micro particles in PWR primary circuit: on-site test

    Itou, A.; Kondo, K.; Kouzuma, Y., E-mail: ayumu_itou@kyuden.co.jp [Kyusyu Electric Power Co., Inc., Minami-ku, Fukuoka (Japan); Umehara, R.; Shimizu, Y., E-mail: Ruyji_Umehara@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., Hyogo-ku, Kobe (Japan); Kogawa, N.; Nagamine, K., E-mail: nkogawa@ndc.hq.mhi.co.jp [Nuclear Development Corp., Tokaimura, Ibaraki (Japan)

    2010-07-01

    In Japanese PWR plants, efforts to remove particulate constituents containing radioactive cobalt which provides a source of radiation exposure, are needed. Performance evaluation study was conducted for macroporous resin which was said to possess excellent performance in removing particulate constituents and whose practical accomplishment at plants in USA was reported to be good. As one of the means for radiation exposure reduction in PWR, a study for application of crud removing resin to actual plant was executed by laboratory experiments using simulated crud (Fe{sub 3}O{sub 4} particle). In this study, following two mechanisms were demonstrated as the particle capturing mechanism of macroporous resin; physical trapping by fine pores on resin surface; electrical adsorption onto resin surface. In addition, in parallel to the study for application of macroporous resin to actual PWR plant, on-site study was planned to investigate the primary system water chemistry during various stages of actual plant operation and to research performance of particle capturing in detail. As the on-site study, column experiments, there water was let pass through the column, were planned for various operation stage (startup period, power operation period and shutdown period). A kind of conventional gel-type resin and three kinds of macroporous resin were examined for onsite tests. As to particulate capturing, basic knowledge regarding capturing efficiency and influence of water chemistry on capturing performance were ordered. Capturing performance of each resin tested became clear and was ordered by comparison. Effectiveness of macroporous resin with regard to crud removal in primary coolant was confirmed. (author)

  18. Containment fan cooler heat transfer calculation during main steam line break for Maanshan PWR plant

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Kao, Lain-Su, E-mail: lskao@iner.gov.tw

    2013-10-15

    Highlights: • Evaluate component cooling water (CCW) thermal response during MSLB for Maanshan. • Using GOTHIC to calculate CCW temperature and determine time required to boil CCW. • Both convective and condensation heat transfer from the air side are considered. • Boiling will not occur since T{sub B} is sufficiently longer than CCW pump restart time. -- Abstract: A thermal analysis has been performed for the Containment Fan Cooler Unit (FCU) during Main Steam Line Break (MSLB) accident, concurrent with loss of offsite power, for Maanshan PWR plant. The analysis is performed in order to address the waterhammer and two-phase flow issues discussed in USNRC's Generic Letter 96-06 (GL 96-06). Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt for each unit. The design basis for containment temperature is Main Steam Line Break (MSLB) accident at power of 2830.5 MWt, which results in peak vapor temperature of 387.6 °F. The design is such that when MSLB occurs concurrent with loss of offsite power (MSLB/LOOP), both the coolant pump on the secondary side and the fan on the air side of the FCU loose power and coast down. The pump has little inertia and coasts down in 2–3 s, while the FCU fan coasts down over much longer period. Before the pump is restored through emergency diesel generator, there is potential for boiling the coolant in the cooling coils by the high-temperature air/steam mixture entering the FCU. The time to boiling depends on the operating pressure of the coolant before the pump is restored. The prediction of the time to boiling is important because it determines whether there is potential for waterhammer or two-phase flow to occur before the pump is restored. If boiling occurs then there exists steam region in the pipe, which may cause the so called condensation induced waterhammer or column closure waterhammer. In either case, a great amount of effort has to be spent to

  19. Valve inlet fluid conditions for pressurizer safety and relief valves for B and W 177-FA and 205-FA plants. Final report. [PWR

    Cartin, L.R.; Winks, R.W.; Merchent, J.W.; Brandt, R.T.

    1982-12-01

    The overpressurization transients for the Babcock and Wilcox Company's 177- and 205-FA units are reviewed to determine the range of fluid conditions expected at the inlet of pressurizer safety and relief valves. The final Safety Analysis Report, extended high-pressure injection, and cold overpressurization events are considered. The results of this review, presented in the form of tables and graphs, provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI PWR Safety and Relief Valve Test Program are representative of those expected in their unit(s).

  20. A MATLAB-Linked Solver to Find Fuel Depletion in a PWR, a Suggested VVER-1000 Type

    F. Faghihi

    2009-01-01

    Full Text Available Coupled first-order IVPs are frequently used in many parts of engineering and sciences. We present a “solver” including three computer programs which were joint with the MATLAB software to solve and plot solutions of the first-order coupled stiff or nonstiff IVPs. Some applications related to IVPs are given here using our MATLAB-linked solver. Muon catalyzed fusion in a D-T mixture is considered as a first dynamical example of the coupled IVPs. Then, we have focused on the fuel depletion in a suggested PWR including poisons burnups (xenon-135 and samarium-149, plutonium isotopes production, and uranium depletion.

  1. Severe accident modeling of a PWR core with different cladding materials

    Johnson, S. C. [Westinghouse Electric Company LLC, 5801 Bluff Road, Columbia, SC 29209 (United States); Henry, R. E.; Paik, C. Y. [Fauske and Associates, Inc., 16W070 83rd Street, Burr Ridge, IL 60527 (United States)

    2012-07-01

    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

  2. Performance evaluation of PSO and GA in PWR core loading pattern optimization

    Khoshahval, F., E-mail: f_khoshahval@sbu.ac.i [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of); Minuchehr, H. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of); Zolfaghari, A., E-mail: a-zolfaghari@sbu.ac.i [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of)

    2011-03-15

    Research highlights: The performance of both GA and PSO methods in optimizing of a PWR core are adequate. It seems GA arrives to its final parameter value in a fewer generation than the PSO. The computation time for GA is higher than PSO. The GA-2 and PSO-CFA algorithms perform better in comparison to GA-1 and PSO-IWA. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Recently, genetic algorithm (GA) and particle swarm optimization (PSO) techniques have attracted considerable attention among various modern heuristic optimization techniques. GA is a powerful optimization technique, based upon the principles of natural selection and species evolution. GA is finding popularity as design tools because of its versatility, intuitiveness and ability to solve highly non-linear, mixed integer optimization problems. PSO refers to a relatively new family of algorithms and is mainly inspired by social behavior patterns of organisms that live within large group. This study addresses the application and performance comparison of PSO and GA optimization methods for nuclear fuel loading pattern problem. Flattening of power inside the reactor core of Bushehr nuclear power plant (WWER-1000 type) is chosen as an objective function to prove the validity of algorithms. In addition the performance of both optimization techniques in terms of convergence rate and computational time is compared. It is found that, from an evolutionary point of view, the performance of both GA and PSO is quite adequate. But, GA seems to arrive at its final parameter value in a fewer generations than the PSO. It is also noticed that, the computation time for implemented GA in this work is too high in comparison to PSO.

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

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

    2014-12-15

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

  4. On-line PWR RHR pump performance testing following motor and impeller replacement

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  5. PWR core and spent fuel pool analysis using scale and nestle

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  6. VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

    Salko, Robert K [ORNL; Sung, Yixing [Westinghouse Electric Company, Cranberry Township; Kucukboyaci, Vefa [Westinghouse Electric Company, Cranberry Township; Xu, Yiban [Westinghouse Electric Company, Cranberry Township; Cao, Liping [Westinghouse Electric Company, Cranberry Township

    2016-01-01

    The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

  7. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  8. Reference neutron transport calculation note for Korea nuclear power plants with 3-loop PWR reactors

    Kim, Byung Cheol; Chang, Ki Oak

    1997-05-01

    Reactor pressure vessel (RPV) steels are subjected to neutron irradiation at a temperature of about 290 deg C. This radiation exposure alters the mechanical properties, leading to a shift of the brittle-to-ductile transition temperature toward higher temperatures and to a diminution of the rupture energy as determined by Charpy V-notch tests. This radiation embrittlement is one of the important aging factors of nuclear power plants. U.S. NRC recommended the basic requirements for the determination of the pressure vessel fluence by regulatory guide DG-1025 in order to reduce the uncertainty in the determination of neutron fluence calculation and measurements. The determination of the pressure vessel fluence is based on both calculations and measurements. The fluence prediction is made with a calculation and the measurements are used to qualify the calculational methodology. Because of the importance and the difficulty of these calculations, the method`s qualification by comparison to measurement must be made to ensure a reliable and accurate vessel fluence determination. This reference calculation note is to provide a series of forward and adjoint neutron transport calculations for use in the evaluation of neutron dosimetry from surveillance capsule irradiations at 3-loop PWR reactor as well as for use in the determination of the neutron exposure of the reactor vessel wall in accordance with U.S Regulatory Guide DG-1025 requirements. The calculations of the pressure vessel fluence consist of the following steps; (1) Determination of the geometrical and material input data, (2) Determination of the core neutron source, and (3) Propagation of the neutron fluence from the core to the vessel and into the cavity. (author). 12 tabs., 3 figs., 7 refs.

  9. Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

    Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier

    2014-06-01

    As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard

  10. Solidification of spent ion exchange resins into the SIAL matrix at the Dukovany NPP, Czech Republic

    Tatransky, Peter; Prazska, Milena; Harvan, David [AMEC Nuclear Slovakia, Trnava, Slovak Republic, 917 01 (Slovakia)

    2013-07-01

    Based on the decision of the State Office for Nuclear Safety, the Dukovany NPP has been obliged to secure the efficient capacities for the disposal of spent ion exchange resins. Therefore, in September 2010, based on the contract with supplier company AMEC Nuclear Slovakia s.r.o. has begun with pumping and treatment of ion exchange resins from the storage tank 0TW30B02, situated in the auxiliary building. The SIAL{sup R} technology, developed in AMEC Nuclear Slovakia, has been used for the solidification purposes. This technology allows an on-site treatment of various special radioactive waste streams (resins, sludge, sludge/resins and borates) at the room temperature. The SIAL{sup R} matrix and technology were licensed by the Czech State Office for Nuclear Safety in 2007. On-site treatment and solidification of spent ion exchange resins at Dukovany NPP involves process of resin removal from tank using remotely operated manipulator, resin transportation, resin separation from free water, resin filling into 200 dm{sup 3} drums and solidification into SIAL{sup R} matrix in 200 dm{sup 3} drums using the FIZA S 200 facility. The final product is observed for compressive strength, leachability, radionuclide composition, dose rate, solids and total weight. After meeting the requirements for final disposal and consolidation, the drums are being transported for the final disposal to the Repository at Dukovany site. During the 3 month's trial operation in 2010, and the normal operation in 2011 and 2012, 189 tons of dewatered resins have been treated into 1960 drums, with total activity higher than 920 GBq. At the end of trial run (2010), 22 tons of dewatered resins were treated into 235 drums. During standard operation approximately 91 tons in 960 drums (2011) and 76 tons in 765 drums (2012) were treated. The weights of resins in the drum ware in the range from 89 - 106 kg and compressive strength limit (10 MPa) has already been achieved 24 hours after fixation. The

  11. Modeling the transport of nitrogen in an NPP-2006 reactor circuit

    Stepanov, O. E.; Galkin, I. Yu.; Sledkov, R. M.; Melekh, S. S.; Strebnev, N. A.

    2016-07-01

    Efficient radiation protection of the public and personnel requires detecting an accident-initiating event quickly. Specifically, if a heat-exchange tube in a steam generator is ruptured, the 16N radioactive nitrogen isotope, which contributes to a sharp increase in the steam activity before the turbine, may serve as the signaling component. This isotope is produced in the core coolant and is transported along the circulation circuit. The aim of the present study was to model the transport of 16N in the primary and the secondary circuits of a VVER-1000 reactor facility (RF) under nominal operation conditions. KORSAR/GP and RELAP5/Mod.3.2 codes were used to perform the calculations. Computational models incorporating the major components of the primary and the secondary circuits of an NPP-2006 RF were constructed. These computational models were subjected to cross-verification, and the calculation results were compared to the experimental data on the distribution of the void fraction over the steam generator height. The models were proven to be valid. It was found that the time of nitrogen transport from the core to the heat-exchange tube leak was no longer than 1 s under RF operation at a power level of 100% N nom with all primary circuit pumps activated. The time of nitrogen transport from the leak to the γ-radiation detection unit under the same operating conditions was no longer than 9 s, and the nitrogen concentration in steam was no less than 1.4% (by mass) of its concentration at the reactor outlet. These values were obtained using conservative approaches to estimating the leak flow and the transport time, but the radioactive decay of nitrogen was not taken into account. Further research concerned with the calculation of thermohydraulic processes should be focused on modeling the transport of nitrogen under RF operation with some primary circuit pumps deactivated.

  12. Development of regulatory guide for review of aging management of the operating NPP

    Shin, Tae Myung; Lee, Jae Kyung [Cheongju Univ., Cheongju (Korea, Republic of); Kim, Young Ryul [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2001-03-15

    This is the final report of the second year study. Based on the first year study, proposal of revised guidelines, analysis of revised or newly issued IAEA safety guides and reference guidelines of developed countries, and proposal of detailed guidelines of aging management in PSR have been performed in the second year study. The summary of results in the study so far can be summarized as below, overall view on PSR and idea of effective domestic application were leaded through additional investigation and comparison of legal basis, experiences and current status of PSR implementation among the countries having operating NPPs including Korea. Strategies of adequate application of PSR are roughly reevaluated and totally reestablished in summary from the analysis in factor by factor basis of PSR implementation experience in foreign countries and background of IAEA guidelines. Models and draft framework of PSR report in the first year study were summarized and reevaluated, and structure and outline options of PSR guidelines for judging the PSR report are newly proposed with comparison of their strengths and weaknesses based on the first year study. Among the opt ions, guidelines framework equivalent to the PSR report was picked up as the best. For the judgement of aging management, the most appropriate one was chosen for the detailed judgement of aging management review in our PSR being based on the Standard Review Plan for License Renewal (SRP-LR) in United States considering potential future usage in the judgement for continued operation of old NPP at the time of expiration of its design life. A draft PSR guidelines is prepared and attached by revision of basic guidelines issued in 2000, considering the issues discussed for the draft revision of IAEA PSR guide, the draft IAEA document about 'experience of PSR implementation of member states', and the characteristics of Hungarian PSR Guidelines.

  13. Environmental qualification design for NPP refurbishment to comply with revised licensing requirements

    MacBeth, M. J.; Hemmings, R. L. [Canatom-NPM, Ontario (Canada)

    2002-04-15

    Recent Canadian Nuclear Regulatory decisions have imposed Environmental Qualification (EQ) requirements for twenty-four Reactor Building (RB) airlocks at the four-unit Pickering Nuclear Generating Station-B (PNGS-B) facility. This paper describes the EQ modification design work completed by CANATOM-NPM for the problematic aspects for such projects. The airlocks allow RB access while providing a containment boundary and are designed to prevent a potential breach of containment for all analysed station conditions. Each PNGS-B unit has three large equipment airlocks and three smaller personnel airlocks. The airlocks must function under postulated worst-case design basis accident(DBA) conditions for assigned mission durations. The design must ensure that accident conditions cannot spuriously initiate an un-requested door opening. CANATOM-NPM reviewed site data to specify the necessary EQ modifications required to satisfy licensing requirements while providing a correct and complete as-found record of the existing airlock installation. The design team assessed the installed airlocks configuration against environmental qualification requirements to finalize the list of necessary modifications. A comprehensive, cross-discipline review of proposed design changes was completed to identify any further changes required to satisfy the final EQ licensing goal. The design team also conducted a design review of the EQ modification installation strategy to integrate the design deliverables with the installation team requirements while attempting to minimize necessary outage time for EQ modification installations. This project was completed on schedule and within the cost limitations required by the client with comprehensive, high quality final design packages. Overall improvements were realized for OPG system drawings and the electronic documentation of design data. The EQ modifications designed by CANATOM-NPM will ensure the continued operation of the PNGS-B NPP past December 31

  14. S-NPP VIIRS instrument telemetry and calibration data trend study

    Sun, ZiPing; De Luccia, Frank J.; Cardema, Jason C.; Moy, Gabriel

    2015-09-01

    The Suomi National Polar Orbiting Partnership (S-NPP) Visible Infrared Imaging Radiometer Suite (VIIRS) employs a large number of temperature and voltage sensors (telemetry points) to monitor instrument health and performance. We have collected data and built tools to study telemetry and calibration parameters trends. The telemetry points are organized into groups based on locations and functionalities. Examples of the groups are: telescope motor, focal plane array (FPA), scan cavity bulkhead, radiators, solar diffuser and Solar Diffuser Stability Monitor (SDSM). We have performed daily monitoring and long-term trending studies. Daily monitoring processes are automated with alarms built into the software to indicate if pre-defined limits are exceeded. Long-term trending studies focus on instrument performance and sensitivities of Sensor Data Record (SDR) products and calibration look-up tables (LUTs) to instrument temperature and voltage variations. VIIRS uses a DC Restore (DCR) process to periodically correct the analog offsets of each detector of each spectral band to ensure that the FPA output signals are always within the dynamic range of the Analog to Digital Converter (ADC). The offset values are updated based on observations of the On-Board Calibrator Blackbody source. We have performed a long-term trend study of DCR offsets and calibration parameters to explore connections of the DCR offsets with onboard calibrators. The study also shows how the instrument and calibration parameters respond to the VIIRS Petulant Mode, spacecraft (SC) anomalies and flight software (FSW) updates. We have also shown that trending studies of telemetry and calibration parameters may help to improve the instrument calibration processes and SDR Quality Flags.

  15. Validation of the Suomi NPP VIIRS Ice Surface Temperature Environmental Data Record

    Yinghui Liu

    2015-12-01

    Full Text Available Continuous monitoring of the surface temperature is critical to understanding and forecasting Arctic climate change; as surface temperature integrates changes in the surface energy budget. The sea-ice surface temperature (IST has been measured with optical and thermal infrared sensors for many years. With the IST Environmental Data Record (EDR available from the Visible Infrared Imaging Radiometer Suite (VIIRS onboard the Suomi National Polar-orbiting Partnership (NPP and future Joint Polar Satellite System (JPSS satellites; we can continue to monitor and investigate Arctic climate change. This work examines the quality of the VIIRS IST EDR. Validation is performed through comparisons with multiple datasets; including NASA IceBridge measurements; air temperature from Arctic drifting ice buoys; Moderate Resolution Imaging Spectroradiometer (MODIS IST; MODIS IST simultaneous nadir overpass (SNO; and surface air temperature from the National Centers for Environmental Prediction/National Center for Atmospheric Research (NCEP/NCAR reanalysis. Results show biases of −0.34; −0.12; 0.16; −3.20; and −3.41 K compared to an aircraft-mounted downward-looking pyrometer; MODIS; MODIS SNO; drifting buoy; and NCEP/NCAR reanalysis; respectively; root-mean-square errors of 0.98; 1.02; 0.95; 4.89; and 6.94 K; and root-mean-square errors with the bias removed of 0.92; 1.01; 0.94; 3.70; and 6.04 K. Based on the IceBridge and MODIS results; the VIIRS IST uncertainty (RMSE meets or exceeds the JPSS system requirement of 1.0 K. The product can therefore be considered useful for meteorological and climatological applications.

  16. Development of a light weighted mobile robot for SG tube inspection in NPP

    Seo, Yong Chil; Jeong, Kyung Min; Shin, Hochul; Gweng, Jung Ju; Lee, Sung Uk; Jeong, Seung Ho; Choi, Young Soo; Kim, Seung Ho [KAERI, Daejeon (Korea, Republic of); Shin, Chun Sup; Park, Ki Tae [Korea Plant Service and Engineering, Busan (Korea, Republic of)

    2012-10-15

    Steam generators (SG) are among the most critical components of pressurized water Nuclear Power Plants (NPP). SG tubes must provide a reliable pressure boundary between the primary and secondary cooling water, because any leakage from tube defects could result in the release of radioactivity to the environment. Thus degradations of steam generators tubes should be monitored and inspected periodically under nuclear regulation. In service inspections of SG tubes are carried out using eddy current test (ECT) and the defected tubes are usually plugged. Because the radioactivity in the internal SG chambers limits free access of human workers, remote manipulators are required. In South Korea, Manipulators such as the Zet ec SM series and the Westinghouse ROSA series have bee used. Such manipulators are rigidly mounted to man ways or tube sheets of SG. Confusions of the inspected tubes may occur from deflection of the manipulators. To reduce the deflections of the manipulators for covering the large working areas of tube sheets, sufficient rigidity is required and that leads to an increase of the weight. Such weight increase results in some difficulties for handling and more radiation exposure of human workers. Recently light weighed mobile robots have been introduced by Westinghouse and Zet ec. The robots can move keeping in contact with the tube sheets using devices which are commonly called cam locks. They are easier to handle and provide no confusion for the position of the inspected tubes. But when the clamping forces are loosed accidentally, they can be fall down and light repair works can be performed. This paper provides the design results for a lightweight mobile robot which is being developed in cooperation of our institutes.

  17. Assessment of PWR fuel degradation by post-irradiation examinations and modeling in DEGRAD-1 code; Avaliacao da degradacao de combustivel PWR por exames pos-irradiacao e modelagem no codigo DEGRAD-1

    Castanheira, Myrthes; Lucki, Georgi; Silva, Jose Eduardo Rosa da; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear]. E-mail: myrthes@ipen

    2005-07-01

    On the majority of the cases, the inquiries on primary failures and secondary in PWR fuel rods are based on results of analysis were made use of the non-destructive examination results (coolant activities monitoring, sipping tests, visual examination). The complementary analysis methodology proposed in this work includes a modeling approach to characterization of the physical effects of the individual chemistry mechanisms that constitute the incubation phase of degradation phenomenon after primary failure that are integrated in the reactor operational history under stationary operational regime, and normal power transients. The computational program called DEGRAD-1 was developed based on this modeling approach. The practical outcome of the program is to predict cladding regions susceptible to massive hydriding. The applications presented demonstrate the validity of proposed method and models by actual cases simulation, which (primary and secondary) defects positions were known and formation time was estimated. By using the modeling approach, a relationship between the hydrogen concentration in the gap and the inner cladding oxide thickness has been identified which, when satisfied, will induce massive hydriding. The novelty in this work is the integrated methodology, which supplements the traditional analysis methods (using data from non-destructive techniques) with mathematical models for the hydrogen evolution, oxidation and hydriding that include refined approaches and criteria for PWR fuel, and using the FRAPCON-3 fuel performance code as the basic tool. (author)

  18. Heat-transfer analysis of double-pipe heat exchangers for indirect-cycle SCW NPP

    Thind, Harwinder

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 °C. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option for PT and PV SCWRs without introducing steam-reheat channels into a reactor. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, steam generators separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in a reactor containment building. This study analyzes the heat transfer from a SCW primary (reactor) loop to a SCW and Super-Heated Steam (SHS) secondary (turbine) loop using a double-pipe intermediate HX. The numerical model is developed with MATLAB and NIST REFPROP software. Water from the primary loop flows through the inner pipe, and water from the secondary loop flows through the annulus in the counter direction of the double-pipe HX. The analysis on the double-pipe HX shows temperature and profiles of thermophysical properties along the heated length of the HX. It was found that the pseudocritical region has a significant effect on the temperature profiles and heat-transfer area of the HX. An analysis shows the effect of variation in pressure, temperature, mass flow rate, and pipe size on the pseudocritical region and the heat-transfer area of the HX. The results from the numerical model can be used to optimize the heat-transfer area of the HX. The higher pressure difference on the hot side and higher temperature difference between the hot and cold sides reduces the pseudocritical-region length, thus

  19. Retention of PWR primary coolant trace elements by cation exchange resins during cold shutdown with oxygenation: modelling and experimental results for silver behavior; Retention des elements traces du fluide primaire des REP par les resines echangeuses de cations lors des mises en arret a froid avec oxygenation: modelisation et resultats experimentaux relatifs au comportement de l'argent

    Elain, L.; Doury-Berthod, M. [CEA Saclay, INSTN, Institut National des Sciences et Techniques Nucleaires, 91 - Gif-sur-Yvette (France); Genin, J.B. [CEA Cadarache, Dir. de l' Energie Nucleaire (DEN), 13 - Saint-Paul-lez-Durance (France); Berger, M. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France)

    2004-07-01

    In order to minimize the radiochemical impact of the corrosion products on the operation of Pressurized Water Reactors, on-line purification of the primary coolant is carried out. The purification system arranged on the Chemical and Volume Control System is made up of mechanical filters and demineralizers packed with a mixed bed of cation and anion exchange resins. This paper proposes an update on the retention of primary coolant trace elements by the cation exchange resins of the demineralizers during cold shutdowns with oxygenation. The study is first of all devoted to the description of the concentration profiles of the various cation constituents which settle in the demineralizer during purification after oxygenation. For a number of trace elements, localized enrichment zones at the Li{sup +}/Ni(Il) exchange zone are expected to appear in the column. The case of silver is afterwards discussed in detail. Thermodynamic modelling shows that the theoretical retention volume of the metallic element and its degree of enrichment in the column are dependent on the basic composition of the primary coolant and the specific characteristics of the demineralizer cation exchanger. At the Ag{sup +} ion concentration expected in the reactor coolant after oxygenation (between 10{sup -8} mol.L{sup -1} and 10{sup -6} mol.L{sup -1}), the breakthrough of silver should be near-simultaneous with that of nickel. The experimental results, obtained in the laboratory and with a 'Mini-CVCS' pilot instrumentation recently used during the cold shutdown of Tricastin Unit 2,900 MWe PWR NPP, confirm the validity of these theoretical forecasts and enable new hypotheses to be advanced for explaining silver release from a demineralizer. (authors)

  20. INTEGRATION OF NPP SEMI MECHANISTIC - MODELLING, REMOTE SENSING AND CIS IN ESTIMATING CO 2 ABSORPTION OF FOREST VEGETATION IN LORE LINDU NATIONAL PARK

    GODE GRAVENHORsr

    2006-01-01

    Full Text Available Net Primary Production, NPP, is one of the most important variables characterizing the performance of an ecosystem. It is the difference between the total carbon uptake from the air through photosynthesis and the carbon loss due to respiration by living plants. However, field measurements of NPP are time-consuming and expensive. Current techniques are therefore not useful for obtaining NPP estimates over large areas. By combining the remote sensing and GIS technology and modelling, we can estimate NPP of a large ecosystem with a little ease. This paper discusses the use of a process based physiological sunshade canopy models in estimating NPP of Lore Lindu National Park (LLNP. The discussion includes on how to parameterize the models and how to scale up from leaf to the canopy. The version documented in this manuscript is called NetPro Model, which is a potential NPP model where water effect is not included yet. The model integrates CIS and the use of Remote Sensing, and written in Visual Basic 6.0 programming language and Map Objects 2.1. NetPro has the capability of estimating NPP of Cs vegetation under present environmental condition and under future scenarios (increasing [CO2], increasing temperature and increasing or decreasing leaf nitrogen level. Based on site-measured parameterisation of VaM* (Photosynthetic capacity, /Jj (Respiration and leaf nitrogen ONi, the model was run under increasing CO2 level and temperature and varied leaf nitrogen. The output of the semi-mechanistic modelling is radiation use efficiency (?. Analysis of remote sensing data give Normalized Difference Vegetation Index (NDVI and related Leaf Area Index (LAI and traction of absorbed Photosynthetically Active Radiation (/M > AK. Climate data are obtained from 12 meteorological stations around die parks, which includes global radiations, minimum and maximum temperature. CO2 absorbed by vegetation (Gross Primary Production, GPP is then calculated using the above

  1. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Andrej Prošek

    2013-01-01

    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

  2. Overview and Discussion of the OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests

    M. Avramova

    2013-01-01

    Full Text Available The Pennsylvania State University (PSU under the sponsorship of the US Nuclear Regulatory Commission (NRC has prepared, organized, conducted, and summarized the Organisation for Economic Co-operation and Development/US Nuclear Regulatory Commission (OECD/NRC benchmark based on the Nuclear Power Engineering Corporation (NUPEC pressurized water reactor (PWR subchannel and bundle tests (PSBTs. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency (NEA of OECD and the Japan Nuclear Energy Safety Organization (JNES, Japan. The OECD/NRC PSBT benchmark was organized to provide a test bed for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational fluid dynamics (CFDs codes. The benchmark was designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and department from nucleate boiling (DNB, under steady-state and transient conditions, to full-scale experimental data. This paper provides an overview of the objectives of the benchmark along with a definition of the benchmark phases and exercises. The NUPEC PWR PSBT facility and the specific methods used in the void distribution measurements are discussed followed by a summary of comparative analyses of submitted final results for the exercises of the two benchmark phases.

  3. The Effects of Hot Bending on the Low Cycle Fatigue Behaviors of 347 SS in PWR Primary Environment

    Kim, Ho-Sub; Hong, Jong-Dae; Lee, Junho; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    Fatigue damage could be significant for some locations, especially the welds and bends where stress concentration is typically high. As a possible solution, a large radius hot-bending method has been suggested to eliminate some weld joints and all tight bends. However, for the hot-bending process which involves a high temperature thermal cycle, there is a concern about changes in mechanical properties including low cycle fatigue behaviors. In APR1400, Type 347 SS have been used as surge line pipes. Therefore, to verify the applicability of hot-bending on 347 SS surge line pipes, an environmental fatigue test program was initiated. In this paper, the preliminary results of the on-going test program are introduced. Also, the low cycle fatigue behaviors of 347 SS are compared with those of other grade of stainless steels. The effects of hot bending on the low cycle fatigue behavior of 347 SS were quantitatively evaluated. The fatigue life was compared with the estimated values per NUREG 6909 rev. 1. There are no distinct differences between NUREG 6909 and LCF tests. According to fractography and cross section analysis in progress, basically, the reduction of LCF life of 347 SS in PWR water was caused by operation of HIC mechanism. The cyclic stress responses shows that there is no secondary hardening in 330 .deg.C air and PWR water.

  4. Analysis of a bending test on a full-scale PWR hot leg elbow containing a surface crack

    Delliou, P. le [Electricite de France, EDF, 77 - Moret-sur-Loing (France). Dept. MTC; Julisch, P.; Hippelein, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Bezdikian, G. [Electricite de France, EDF, 92 - Paris la Defense (France). Direction Production Transport

    1998-11-01

    EDF, in co-operation with Framatome, has conducted a large research programme on the mechanical behaviour of thermally aged cast duplex stainless steel elbows, which are part of the main primary circuit of French PWR. One important task of this programme consisted of testing a full-scale PWR hot leg elbow. The elbow contained a semi-elliptical circumferential notch machined on the outer surface of the intrados as well as casting defects located on the flanks. To simulate the end-of-life condition of the component regarding material toughness, it had undergone a 2400 hours ageing heat treatment at 400 C. The test preparation and execution, as well as the material characterization programme, were committed to MPA. The test was conducted under constant internal pressure and in-plane bending (opening mode) at 200 C. For safety reasons, it took place on an open air-site: the Meppen military test ground. At the maximum applied moment (6000 kN.m), the notch did not initiate. This paper presents the experimental results and the fracture mechanics analysis of the test, based on finite element calculations. (orig.)

  5. Improvement of availability of PWR nuclear plants through the reduction of the time required for refueling/maintenance outages

    Mayers, J.B.; Soth, L.G.

    1978-04-01

    The objective of the project, conducted by Commonwealth Research Corporation and Westinghouse Electric Corporation, is to identify improvements in procedures and equipment which will reduce the time required for refueling/maintenance outages at PWR nuclear power plants. The outage of Commonwealth Edison Zion Station Unit 1 in March through May of 1976 was evaluated to identify those items which caused delays and those work activities that offer the potential for significant improvements that could reduce the overall duration of the outage and achieve an improvement in the plant's availability for power production. Modifications in procedures have been developed and were evaluated during one or more outages in 1977. Conceptual designs have been developed for equipment modifications to the refueling system that could reduce the time required for the refueling portion of the outage. The purpose of the interim report is to describe those conceptual designs and to assess their impact upon future outages. Recommendations are included for the implementation of these equipment improvements in a continuation of this program as a demonstration of plant availability benefits that can be realized in PWR nuclear plants already in operation or under construction.

  6. Development of CHF correlation “MG-NV” for low pressure and low velocity conditions applied to PWR safety analysis

    Yumura, T.; Yodo, T.; Makino, Y.; Suemura, T. [Mitsubishi Heavy Industries, LTD., Kobe, Hyogo (Japan)

    2011-07-01

    The Critical Heat Flux (CHF) is one of the important parameters in the safety analysis of Pressurized Water Reactor (PWR). If the CHF is reached, an abrupt drop occurs in the heat transfer between the fuel rod cladding and the reactor coolant, which may induce a large temperature excursion of fuel cladding and a subsequent fuel failure. Therefore, accurate prediction of CHF is required in order to assure a sufficient safety margin in the PWR core. Mitsubishi Heavy Industries, ltd (MHI) is developing a new series of CHF correlations which covers various fuel designs and wide range of fluid conditions with sufficient reliability. In this paper, a new CHF correlation, MG-NV (Mitsubishi Generalized correlation for Non-Vane grid spacers) is presented. This correlation is one of the basic components of the new correlation series and was developed to cover low pressure and low velocity conditions where the rod bundle CHF data are limited. The CHF correlation was developed based on open CHF database and provides conservative but more reliable rod bundle CHF predictions compared with the conventional CHF correlations used in safety analyses at low pressure condition, such as Main Steam Line Break event. (author)

  7. Surface Water Mapping from Suomi NPP-VIIRS Imagery at 30 m Resolution via Blending with Landsat Data

    Chang Huang

    2016-07-01

    Full Text Available Monitoring the dynamics of surface water using remotely sensed data generally requires both high spatial and high temporal resolutions. One effective and popular approach for achieving this is image fusion. This study adopts a widely accepted fusion model, the Enhanced Spatial and Temporal Adaptive Reflectance Fusion Model (ESTARFM, for blending the newly available coarse-resolution Suomi NPP-VIIRS data with Landsat data in order to derive water maps at 30 m resolution. The Pan-sharpening technique was applied to preprocessing NPP-VIIRS data to achieve a higher-resolution before blending. The modified Normalized Difference Water Index (mNDWI was employed for mapping surface water area. Two fusion alternatives, blend-then-index (BI or index-then-blend (IB, were comparatively analyzed against a Landsat derived water map. A case study of mapping Poyang Lake in China, where water distribution pattern is complex and the water body changes frequently and drastically, was conducted. It has been revealed that the IB method derives more accurate results with less computation time than the BI method. The BI method generally underestimates water distribution, especially when the water area expands radically. The study has demonstrated the feasibility of blending NPP-VIIRS with Landsat for achieving surface water mapping at both high spatial and high temporal resolutions. It suggests that IB is superior to BI for water mapping in terms of efficiency and accuracy. The finding of this study also has important reference values for other blending works, such as image blending for vegetation cover monitoring.

  8. Measurement of long-lived radionuclides in surface soil around F1NPP accident site by Accelerator Mass Spectrometry

    Miyake, Yasuto; Matsuzaki, Hiroyuki; Sasa, Kimikazu; Takahashi, Tsutomu

    2015-10-01

    In March 2011, vast amounts of radionuclides were released into the environment due to the Fukushima Daiichi Nuclear Power Plant (F1NPP) accident. However, very little work has been done concerning accident-derived long-lived nuclides such as 129I (T1/2 = 1.57 × 107 year) and 36Cl (T1/2 = 3.01 × 105 year). 129I and 131I are both produced by 235U fission in nuclear reactors. Being isotopes of iodine, these nuclides are expected to behave similarly in the environment. This makes 129I useful for retrospective reconstruction of 131I distribution during the initial stages of the accident. On the other hand, 36Cl is generated during reactor operation via neutron capture reaction of 35Cl, an impurity in the coolant or reactor component. Resulting 36Cl/Cl ratio within the reactor is thus much higher compared to that in environment. Similar to 129I, 36Cl is expected to have leaked out during the accident and it is important to evaluate its effects. In this study, 129I concentrations were determined in several surface soil samples collected around F1NPP. Average 129I/131I ratio was estimated to be 26.1 ± 5.8 as of March 11, 2011, consistent with calculations using ORIGEN2 code and other published data. 36Cl/Cl ratios in some of the soil samples were likewise measured and ranged from 1.1 × 10-12 to 2.6 × 10-11. These are higher compared to ratios measured around F1NPP before the accident. A positive correlation between 36Cl and 129I concentration was observed.

  9. Training in fundamentals of radiological coverage in Laguna Verde NPP; Entrenamiento de fundamentos de coberturas radiologicas en la CLV

    Lara H, M. A., E-mail: marcolarah@gmail.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    In 2010, the Institute of Nuclear Power Operations (INPO) celebrates the Knowledge Transfer and Retention Workshop, an event where nuclear regulators and operators presented the strategies that various NPP to worldwide were implemented to mitigate the consequences of this generational change and take advantage of it, the trend in the presented works was the same: the generational change occurs in a faster way that the transfer of knowledge, the future was already here and many NPP had not been adequately prepared to train its nuclear technicians and engineers in the tasks demanded by the industry of them, so in addition to preparing these workers to forced marches was necessary to establish strategies to retain at more experienced staff in the industry. The Laguna Verde NPP has not been exempt to this process; the preparation of personnel squares to replace those that reaching retirement age in the Comision Federal de Electricidad (CFE) has become extensive in the last five years, sometimes leading to have personnel covering functions without an alternate to the next lower position, the cause? Not enough staff. In the specific case of radiation protection (Rp) the time required for obtaining the status of Rp technician according to the ANSI/ANS 3.1 standard is 2 years, one of the tasks that most occupies part of these two years is training in radiological coverage, this training requires a mix of knowledge and experience, recently one of the concepts used for training in Rp is the evaluation and management of the radiological risk, topic that is considered in this technical work. (Author)

  10. Measurement of long-lived radionuclides in surface soil around F1NPP accident site by Accelerator Mass Spectrometry

    Miyake, Yasuto; Matsuzaki, Hiroyuki [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo (Japan); Sasa, Kimikazu; Takahashi, Tsutomu [AMS Group, Tandem Accelerator Complex, Research Facility Center for Science and Technology, University of Tsukuba, Tsukuba, Ibaraki 305-8577 (Japan)

    2015-10-15

    In March 2011, vast amounts of radionuclides were released into the environment due to the Fukushima Daiichi Nuclear Power Plant (F1NPP) accident. However, very little work has been done concerning accident-derived long-lived nuclides such as {sup 129}I (T{sub 1/2} = 1.57 × 10{sup 7} year) and {sup 36}Cl (T{sub 1/2} = 3.01 × 10{sup 5} year). {sup 129}I and {sup 131}I are both produced by {sup 235}U fission in nuclear reactors. Being isotopes of iodine, these nuclides are expected to behave similarly in the environment. This makes {sup 129}I useful for retrospective reconstruction of {sup 131}I distribution during the initial stages of the accident. On the other hand, {sup 36}Cl is generated during reactor operation via neutron capture reaction of {sup 35}Cl, an impurity in the coolant or reactor component. Resulting {sup 36}Cl/Cl ratio within the reactor is thus much higher compared to that in environment. Similar to {sup 129}I, {sup 36}Cl is expected to have leaked out during the accident and it is important to evaluate its effects. In this study, {sup 129}I concentrations were determined in several surface soil samples collected around F1NPP. Average {sup 129}I/{sup 131}I ratio was estimated to be 26.1 ± 5.8 as of March 11, 2011, consistent with calculations using ORIGEN2 code and other published data. {sup 36}Cl/Cl ratios in some of the soil samples were likewise measured and ranged from 1.1 × 10{sup −12} to 2.6 × 10{sup −11}. These are higher compared to ratios measured around F1NPP before the accident. A positive correlation between {sup 36}Cl and {sup 129}I concentration was observed.

  11. The challenge of the global management of plant design modifications. example of the new EJ system at Vandellos NPP

    Ortega, Fernando; Valdivia, Carlos; Fernandez Illobre, Luis; Trueba, Pedro [Control Rooms and Simulation, Tecnatom, Avda. Montes de Oca, 1 - 28703 San Sebastian de los Reyes. Madrid (Spain)

    2010-07-01

    One of the most challenging areas in the operation of nuclear power plants (NPP) is related to the management of plant design modifications. Plant modifications can be made to improve reliability, facilitate operation, improve safety or get better results. In any of these situations, plant modifications imply many different activities that have to be done in a coordinated manner. NUREG-0711 (Human Factors Engineering Program Review Model) shows a global approach to manage most of these activities. Although this approach is mainly focused on the design and construction of new plants, it can also be applied to plant modification management. Successful global management will require performing every activity in a specific order, taking advantage of the output coming from some tasks as input for others and finalizing every task when necessary. This will provide the best results in terms of quality, time required for implementation, safe and reliable operation and maintenance, and cost. Tecnatom is involved in most of the activities related to the operational areas and has applied a global approach to get advantages in terms of quality and cost, which is outlined in this paper. As an example of this approach, the Vandellos NPP experience is shown in this presentation. Vandellos NPP carried out an important design modification that consists of replacing an old essential service water system with a new one. This was a three-year project that implied the construction of new reservoirs, new buildings, the implementation of new equipment, and new panels in the main control room. This paper shows the way in which all of these activities were performed. (authors)

  12. Application of the leak-before-break concept to the primary circuit piping of the Leningrad NPP

    Eperin, A.P.; Zakharzhevsky, Yu.O.; Arzhaev, A.I. [and others

    1997-04-01

    A two-year Finnish-Russian cooperation program has been initiated in 1995 to demonstrate the applicability of the leak-before-break concept (LBB) to the primary circuit piping of the Leningrad NPP. The program includes J-R curve testing of authentic pipe materials at full operating temperature, screening and computational LBB analyses complying with the USNRC Standard Review Plan 3.6.3, and exchange of LBB-related information with emphasis on NDE. Domestic computer codes are mainly used, and all tests and analyses are independently carried out by each party. The results are believed to apply generally to RBMK type plants of the first generation.

  13. Annual and seasonal variations In the gamma activities in Sava river sediments upstream and downstream of NPP Krsko

    Stipe, Lulic [Rudjer Boskovic Institute, Lab. for radioecology, Zagreb (Croatia)

    2006-07-01

    Results of the five years monitoring of artificial and natural occurring radionuclides in the Sava river sediments are presented. Measurements were conducted as a part of the regular Krsko Nuclear Power Plant radioactivity control and the independent supervisions of the input of radionuclides into larger environment (immission). In order to estimate seasonal variations samples were taken from seven locations (one upstream and five downstream of the Krsko NPP) during four sampling period (seasonal) in each year. Selected radionuclides in the sediment fractiess than 0.5 mm were determined with gamma spectrometer equipped with BE3830 model High Purity Ge detector with 30% relative efficiency. (authors)

  14. The effects of cold rolling orientation and water chemistry on stress corrosion cracking behavior of 316L stainless steel in simulated PWR water environments

    Chen, Junjie; Lu, Zhanpeng; Xiao, Qian; Ru, Xiangkun; Han, Guangdong; Chen, Zhen; Zhou, Bangxin; Shoji, Tetsuo

    2016-04-01

    Stress corrosion cracking behaviors of one-directionally cold rolled 316L stainless steel specimens in T-L and L-T orientations were investigated in hydrogenated and deaerated PWR primary water environments at 310 °C. Transgranular cracking was observed during the in situ pre-cracking procedure and the crack growth rate was almost not affected by the specimen orientation. Locally intergranular stress corrosion cracks were found on the fracture surfaces of specimens in the hydrogenated PWR water. Extensive intergranular stress corrosion cracks were found on the fracture surfaces of specimens in deaerated PWR water. More extensive cracks were found in specimen T-L orientation with a higher crack growth rate than that in the specimen L-T orientation with a lower crack growth rate. Crack branching phenomenon found in specimen L-T orientation in deaerated PWR water was synergistically affected by the applied stress direction as well as the preferential oxidation path along the elongated grain boundaries, and the latter was dominant.

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

    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

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

    None, None

    1998-09-01

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

  17. Example Calculations of In{sub v}essel Steam Explosions for a Prototypical PWR

    Park, Ik Kyu; Hong, Seong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    In this paper, the sample calculation for the in{sub v}essel steam explosions were done by using the MC3D code. The evaluation of the computational code had been done against TROI experiments and the code had been adapted to a PWR ex{sub v}essel steam explosion calculations. MC3D is a code for the calculation of different types of multiphase multi-component flows. It has been built with the fuel-coolant interaction calculations in mind. It is, however, able to calculate very different situations and has a rather wide field of potential applications. MC3D is a set of two fuel-coolant interaction codes with a common numeric solver, one for the premixing phase and one for the explosion phase. In general, the steam explosion simulation with MC3D is being carried out in two steps. In the first step, the distributions of the melt, water, and vapor phases at steam explosion triggering are being calculated with the premixing module. These premixing simulation results present the input for the second step when the escalation and propagation of the steam explosion through the premixture are being calculated with the explosion module. The MC3D premixing model is a six-field application in which the melt is described by three fields. The first one is called 'continuous' and can describe many situations as, e.g., a jet or the melt lying on the bottom of a vessel. The second field corresponds to the droplets issued from the jet fragmentation. This field describes the discontinuous state of the fuel. The third field is optional and describes the fuel fragments issuing from drop fine fragmentation. The remaining three fields are the water, the vapor, and a noncondensable gas. The drop surface area is modeled with a standard interfacial area transport equation. In the explosion model, the continuous phase is not present and only the two fields related to the dispersed fuel are considered

  18. Using Ground Targets to Validate S-NPP VIIRS Day-Night Band Calibration

    Xuexia Chen

    2016-11-01

    Full Text Available In this study, the observations from S-NPP VIIRS Day-Night band (DNB and Moderate resolution bands (M bands of Libya 4 and Dome C over the first four years of the mission are used to assess the DNB low gain calibration stability. The Sensor Data Records produced by NASA Land Product Evaluation and Algorithm Testing Element (PEATE are acquired from nearly nadir overpasses for Libya 4 desert and Dome C snow surfaces. A kernel-driven bidirectional reflectance distribution function (BRDF correction model is used for both Libya 4 and Dome C sites to correct the surface BRDF influence. At both sites, the simulated top-of-atmosphere (TOA DNB reflectances based on SCIAMACHY spectral data are compared with Land PEATE TOA reflectances based on modulated Relative Spectral Response (RSR. In the Libya 4 site, the results indicate a decrease of 1.03% in Land PEATE TOA reflectance and a decrease of 1.01% in SCIAMACHY derived TOA reflectance over the period from April 2012 to January 2016. In the Dome C site, the decreases are 0.29% and 0.14%, respectively. The consistency between SCIAMACHY and Land PEATE data trends is good. The small difference between SCIAMACHY and Land PEATE derived TOA reflectances could be caused by changes in the surface targets, atmosphere status, and on-orbit calibration. The reflectances and radiances of Land PEATE DNB are also compared with matching M bands and the integral M bands based on M4, M5, and M7. The fitting trends of the DNB to integral M bands ratios indicate a 0.75% decrease at the Libya 4 site and a 1.89% decrease at the Dome C site. Part of the difference is due to an insufficient number of sampled bands available within the DNB wavelength range. The above results indicate that the Land PEATE VIIRS DNB product is accurate and stable. The methods used in this study can be used on other satellite instruments to provide quantitative assessments for calibration stability.

  19. Probabilistic Seismic Hazard Assessment for a NPP in the Upper Rhine Graben, France

    Clément, Christophe; Chartier, Thomas; Jomard, Hervé; Baize, Stéphane; Scotti, Oona; Cushing, Edward

    2015-04-01

    The southern part of the Upper Rhine Graben (URG) straddling the border between eastern France and western Germany, presents a relatively important seismic activity for an intraplate area. A magnitude 5 or greater shakes the URG every 25 years and in 1356 a magnitude greater than 6.5 struck the city of Basel. Several potentially active faults have been identified in the area and documented in the French Active Fault Database (web site in construction). These faults are located along the Graben boundaries and also inside the Graben itself, beneath heavily populated areas and critical facilities (including the Fessenheim Nuclear Power Plant). These faults are prone to produce earthquakes with magnitude 6 and above. Published regional models and preliminary geomorphological investigations provided provisional assessment of slip rates for the individual faults (0.1-0.001 mm/a) resulting in recurrence time of 10 000 years or greater for magnitude 6+ earthquakes. Using a fault model, ground motion response spectra are calculated for annual frequencies of exceedance (AFE) ranging from 10-4 to 10-8 per year, typical for design basis and probabilistic safety analyses of NPPs. A logic tree is implemented to evaluate uncertainties in seismic hazard assessment. The choice of ground motion prediction equations (GMPEs) and range of slip rate uncertainty are the main sources of seismic hazard variability at the NPP site. In fact, the hazard for AFE lower than 10-4 is mostly controlled by the potentially active nearby Rhine River fault. Compared with areal source zone models, a fault model localizes the hazard around the active faults and changes the shape of the Uniform Hazard Spectrum at the site. Seismic hazard deaggregations are performed to identify the earthquake scenarios (including magnitude, distance and the number of standard deviations from the median ground motion as predicted by GMPEs) that contribute to the exceedance of spectral acceleration for the different AFE

  20. Evaluation of NPP-VIIRS Nighttime Light Data for Mapping Global Fossil Fuel Combustion CO2 Emissions: A Comparison with DMSP-OLS Nighttime Light Data.

    Ou, Jinpei; Liu, Xiaoping; Li, Xia; Li, Meifang; Li, Wenkai

    2015-01-01

    Recently, the stable light products and radiance calibrated products from Defense Meteorological Satellite Program's (DMSP) Operational Linescan System (OLS) have been useful for mapping global fossil fuel carbon dioxide (CO2) emissions at fine spatial resolution. However, few studies on this subject were conducted with the new-generation nighttime light data from the Visible Infrared Imaging Radiometer Suite (VIIRS) sensor on the Suomi National Polar-orbiting Partnership (NPP) Satellite, which has a higher spatial resolution and a wider radiometric detection range than the traditional DMSP-OLS nighttime light data. Therefore, this study performed the first evaluation of the potential of NPP-VIIRS data in estimating the spatial distributions of global CO2 emissions (excluding power plant emissions). Through a disaggregating model, three global emission maps were then derived from population counts and three different types of nighttime lights data (NPP-VIIRS, the stable light data and radiance calibrated data of DMSP-OLS) for a comparative analysis. The results compared with the reference data of land cover in Beijing, Shanghai and Guangzhou show that the emission areas of map from NPP-VIIRS data have higher spatial consistency of the artificial surfaces and exhibit a more reasonable distribution of CO2 emission than those of other two maps from DMSP-OLS data. Besides, in contrast to two maps from DMSP-OLS data, the emission map from NPP-VIIRS data is closer to the Vulcan inventory and exhibits a better agreement with the actual statistical data of CO2 emissions at the level of sub-administrative units of the United States. This study demonstrates that the NPP-VIIRS data can be a powerful tool for studying the spatial distributions of CO2 emissions, as well as the socioeconomic indicators at multiple scales.

  1. Accounting for land use in life cycle assessment: The value of NPP as a proxy indicator to assess land use impacts on ecosystems.

    Taelman, Sue Ellen; Schaubroeck, Thomas; De Meester, Steven; Boone, Lieselot; Dewulf, Jo

    2016-04-15

    Terrestrial land and its resources are finite, though, for economic and socio-cultural needs of humans, these natural resources are further exploited. It highlights the need to quantify the impact humans possibly have on the environment due to occupation and transformation of land. As a starting point of this paper (1(st) objective), the land use activities, which may be mainly socio-culturally or economically oriented, are identified in addition to the natural land-based processes and stocks and funds that can be altered due to land use. To quantify the possible impact anthropogenic land use can have on the natural environment, linked to a certain product or service, life cycle assessment (LCA) is a tool commonly used. During the last decades, many indicators are developed within the LCA framework in an attempt to evaluate certain environmental impacts of land use. A second objective of this study is to briefly review these indicators and to categorize them according to whether they assess a change in the asset of natural resources for production and consumption or a disturbance of certain ecosystem processes, i.e. ecosystem health. Based on these findings, two enhanced proxy indicators are proposed (3(rd) objective). Both indicators use net primary production (NPP) loss (potential NPP in the absence of humans minus remaining NPP after land use) as a relevant proxy to primarily assess the impact of land use on ecosystem health. As there are two approaches to account for the natural and productive value of the NPP remaining after land use, namely the Human Appropriation of NPP (HANPP) and hemeroby (or naturalness) concepts, two indicators are introduced and the advantages and limitations compared to state-of-the-art NPP-based land use indicators are discussed. Exergy-based spatially differentiated characterization factors (CFs) are calculated for several types of land use (e.g., pasture land, urban land).

  2. Evaluation of NPP-VIIRS Nighttime Light Data for Mapping Global Fossil Fuel Combustion CO2 Emissions: A Comparison with DMSP-OLS Nighttime Light Data.

    Jinpei Ou

    Full Text Available Recently, the stable light products and radiance calibrated products from Defense Meteorological Satellite Program's (DMSP Operational Linescan System (OLS have been useful for mapping global fossil fuel carbon dioxide (CO2 emissions at fine spatial resolution. However, few studies on this subject were conducted with the new-generation nighttime light data from the Visible Infrared Imaging Radiometer Suite (VIIRS sensor on the Suomi National Polar-orbiting Partnership (NPP Satellite, which has a higher spatial resolution and a wider radiometric detection range than the traditional DMSP-OLS nighttime light data. Therefore, this study performed the first evaluation of the potential of NPP-VIIRS data in estimating the spatial distributions of global CO2 emissions (excluding power plant emissions. Through a disaggregating model, three global emission maps were then derived from population counts and three different types of nighttime lights data (NPP-VIIRS, the stable light data and radiance calibrated data of DMSP-OLS for a comparative analysis. The results compared with the reference data of land cover in Beijing, Shanghai and Guangzhou show that the emission areas of map from NPP-VIIRS data have higher spatial consistency of the artificial surfaces and exhibit a more reasonable distribution of CO2 emission than those of other two maps from DMSP-OLS data. Besides, in contrast to two maps from DMSP-OLS data, the emission map from NPP-VIIRS data is closer to the Vulcan inventory and exhibits a better agreement with the actual statistical data of CO2 emissions at the level of sub-administrative units of the United States. This study demonstrates that the NPP-VIIRS data can be a powerful tool for studying the spatial distributions of CO2 emissions, as well as the socioeconomic indicators at multiple scales.

  3. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1. [PWR; BWR

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

  4. Tracking on-orbit stability of the response versus scan angle for the S-NPP VIIRS reflective solar bands

    Wu, Aisheng; Xiong, Xiaoxiong (Jack); Cao, Changyong

    2016-09-01

    Built on strong heritage of the MODIS (Moderate Resolution Imaging Spectroradiometer) sensor, the Visible Infrared Imaging Radiometer Suite (VIIRS) carried on Suomi NPP (National Polar-orbiting Partnership) satellite (http://npp.gsfc.nasa.gov/viirs.html) has been in operation for nearly five fives. The on-board calibration of the VIIRS reflective solar bands (RSB) relies on a solar diffuser (SD) located at a fixed scan angle and a solar diffuser stability monitor (SDSM). The VIIRS response versus scan angle (RVS) was characterized prelaunch in lab ambient conditions and is currently used to determine the on orbit response for all scan angles relative to the SD scan angle. Since the RVS is vitally important to the quality of calibrated level 1B products, it is important to monitor its on-orbit stability. In this study, the RVS stability is examined based on reflectance trends collected from 16-day repeatable orbits over preselected pseudo-invariant desert sites in Northern Africa. These trends cover nearly entire Earth view scan range so that any systematic drifts in the scan angle direction would indicate a change in RVS. This study also compares VIIRS RVS on-orbit stability results with those from Aqua and Terra MODIS over the first four years of mission for a few selected bands, which provides further information on potential VIIRS RVS on-orbit changes.

  5. NPP planning based on analysis of ground vibration caused by collapse of large-scale cooling towers

    Lin, Feng; Ji, Hongkui [Department of Structural Engineering, Tongji University, No. 1239 Siping Road, Shanghai 200092 (China); Gu, Xianglin, E-mail: gxl@tongji.edu.cn [Department of Structural Engineering, Tongji University, No. 1239 Siping Road, Shanghai 200092 (China); Li, Yi [Department of Structural Engineering, Tongji University, No. 1239 Siping Road, Shanghai 200092 (China); Wang, Mingreng; Lin, Tao [East China Electric Power Design Institute Co., Ltd, No. 409 Wuning Road, Shanghai 200063 (China)

    2015-12-15

    Highlights: • New recommendations for NPP planning were addressed taking into account collapse-induced ground vibration. • Critical factors influencing the collapse-induced ground vibration were investigated. • Comprehensive approach was presented to describe the initiation and propagation of collapse-induced disaster. - Abstract: Ground vibration induced by collapse of large-scale cooling towers can detrimentally influence the safe operation of adjacent nuclear-related facilities. To prevent and mitigate these hazards, new planning methods for nuclear power plants (NPPs) were studied considering the influence of these hazards. First, a “cooling tower-soil” model was developed, verified, and used as a numerical means to investigate ground vibration. Afterwards, five critical factors influencing collapse-induced ground vibration were analyzed in-depth. These influencing factors included the height and weight of the towers, accidental loads, soil properties, overlying soil, and isolation trench. Finally, recommendations relating to the control and mitigation of collapse-induced ground vibration in NPP planning were proposed, which addressed five issues, i.e., appropriate spacing between a cooling tower and the nuclear island, control of collapse modes, sitting of a cooling tower and the nuclear island, application of vibration reduction techniques, and the influence of tower collapse on surroundings.

  6. Construction scheduled delay risk assessment by using combined AHP-RII methodology for an international NPP project

    Hossen, Mufazal Muhammed; Kang, Sun Koo; Kim, Jong Hyun [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of)

    2015-04-15

    In this study, Nuclear Power Plant (NPP) construction schedule delay risk assessment methodology is developed and the construction delay risk is assessed for turnkey international NPP projects. Three levels of delay factors were selected through literature review and discussions with nuclear industry experts. A questionnaire survey was conducted on the basis of an analytic hierarchy process (AHP) and Relative Importance Index (RII) methods and the schedule delay risk is assessed qualitatively and quantitatively by severity and frequency of occurrence of delay factors. This study assigns four main delay factors to the first level: main contractor, utility, regulatory authority, and financial and country factor. The second and the third levels are designed with 12 sub-factors and 32 sub-sub-factors, respectively. This study finds the top five most important sub-sub-factors, which are as follows: policy changes, political instability and public intervention; uncompromising regulatory criteria and licensing documents conflicting with existing regulations; robust design document review procedures; redesign due to errors in design and design changes; and worldwide shortage of qualified and experienced nuclear specific equipment manufacturers. The proposed combined AHP-RII methodology is capable of assessing delay risk effectively and efficiently. Decision makers can apply risk informed decision making to avoid unexpected construction delays of NPPs.

  7. The possibilities of applying a risk-oriented approach to the NPP reliability and safety enhancement problem

    Komarov, Yu. A.

    2014-10-01

    An analysis and some generalizations of approaches to risk assessments are presented. Interconnection between different interpretations of the "risk" notion is shown, and the possibility of applying the fuzzy set theory to risk assessments is demonstrated. A generalized formulation of the risk assessment notion is proposed in applying risk-oriented approaches to the problem of enhancing reliability and safety in nuclear power engineering. The solution of problems using the developed risk-oriented approaches aimed at achieving more reliable and safe operation of NPPs is described. The results of studies aimed at determining the need (advisability) to modernize/replace NPP elements and systems are presented together with the results obtained from elaborating the methodical principles of introducing the repair concept based on the equipment technical state. The possibility of reducing the scope of tests and altering the NPP systems maintenance strategy is substantiated using the risk-oriented approach. A probabilistic model for estimating the validity of boric acid concentration measurements is developed.

  8. Conceptual Design of Portable Filtered Air Suction Systems For Prevention of Released Radioactive Gas under Severe Accidents of NPP

    Gu, Beom W.; Choi, Su Y.; Yim, Man S.; Rim, Chun T. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    It becomes evident that severe accidents may occur by unexpected disasters such as tsunami, heavy flood, or terror. Once radioactive material is released from NPP through severe accidents, there are no ways to prevent the released radioactive gas spreading in the air. As a remedy for this problem, the idea on the portable filtered air suction system (PoFASS) for the prevention of released radioactive gas under severe accidents was proposed. In this paper, the conceptual design of a PoFASS focusing on the number of robot fingers and robot arm rods are proposed. In order to design a flexible robot suction nozzle, mathematical models for the gaps which represent the lifted heights of extensible covers for given convex shapes of pipes and for the covered areas are developed. In addition, the system requirements for the design of the robot arms of PoFASS are proposed, which determine the accessible range of leakage points of released radioactive gas. In this paper, the conceptual designs of the flexible robot suction nozzle and robot arm have been conducted. As a result, the minimum number of robot fingers and robot arm rods are defined to be four and three, respectively. For further works, extensible cover designs on the flexible robot suction nozzle and the application of the PoFASS to the inside of NPP should be studied because the radioactive gas may be released from connection pipes between the containment building and auxiliary buildings.

  9. Retrospective study of {sup 14}C concentration in the vicinity of NPP Jaslovské Bohunice using tree rings and the AMS technique

    Ješkovský, Miroslav [CENTA Laboratory, Faculty of Mathematics, Physics and Informatics, Comenius University, 84248 Bratislava (Slovakia); Povinec, Pavel P., E-mail: Povinec@fmph.uniba.sk [CENTA Laboratory, Faculty of Mathematics, Physics and Informatics, Comenius University, 84248 Bratislava (Slovakia); Steier, Peter [VERA Laboratory, Faculty of Physics, University of Vienna, 1090 Vienna (Austria); Šivo, Alexander; Richtáriková, Marta [CENTA Laboratory, Faculty of Mathematics, Physics and Informatics, Comenius University, 84248 Bratislava (Slovakia); Golser, Robin [VERA Laboratory, Faculty of Physics, University of Vienna, 1090 Vienna (Austria)

    2015-10-15

    Atmospheric radiocarbon has been monitored around the Nuclear Power Plant (NPP) Jaslovské Bohunice (Slovakia) using CO{sub 2} absorption in NaOH solution since 1969. In 2012, tree ring samples were collected from Tilia cordata using an increment borer at Žlkovce monitoring station situated close to the Bohunice NPP. Each tree ring was identified and graphite targets were produced for {sup 14}C analysis by accelerator mass spectrometry. The {sup 14}C concentrations obtained from the tree-ring samples have been in a reasonable agreement with the averaged annual {sup 14}C concentrations in atmospheric CO{sub 2}.

  10. Gas-liquid countercurrent two-phase flow in a PWR hot leg: A comprehensive research review

    Deendarlianto, E-mail: deendarlianto@ugm.ac.id [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Department of Mechanical and Industrial Engineering, Faculty of Engineering, Gadjah Mada University, Jalan Grafika No. 2, Yogyakarta 55281 (Indonesia); Hoehne, Thomas; Lucas, Dirk [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Vierow, Karen [Department of Nuclear Engineering Texas A and M University, 129 Zachry Engineering Center, 3133 TAMU College Station, TX 77843-3133 (United States)

    2012-02-15

    Highlights: Black-Right-Pointing-Pointer We review the scientific progress on the CCFL in a PWR hot leg. Black-Right-Pointing-Pointer It includes the experimental data, one-dimensional and CFD models in the open literatures. Black-Right-Pointing-Pointer The weak and strong points of the published works were clarified. Black-Right-Pointing-Pointer The research directions in this field were proposed. - Abstract: Research into gas-liquid countercurrent two-phase flow in a model of pressurized water reactor (PWR) hot leg has been carried out over the last several decades. An extensive experimental data base has been accumulated from these studies, leading to the development of phenomenological correlations and scaling parameters of the countercurrent flow limitation (CCFL). However, most of the proposed correlations apply under a relatively narrow range of conditions, generally limited to the test section conditions and/or geometry. Moreover the development of mechanistic models based on the underlying physical processes has been limited. In contrast to this mechanistic form of modelling, the implementation of computational fluid dynamics (CFD) techniques has also been pursued, but the considerable robust three-dimensional (3D) closure relations for this application remain an unachieved goal due to lack of detailed phenomenological knowledge and consequent application of empirical one-dimensional experimental correlations to the multidimensional problem. This paper presents a comprehensive review of research work on countercurrent gas-liquid two-phase flow in a PWR hot leg and provides direction regarding future research on this topic. In the introductory section, the problems facing current research are described. In the following sections, recent experimental as well as theoretical research achievements are overviewed. In the last section, the problems that remain unsolved are discussed, along with some concluding remarks. It was found that only limited theoretical

  11. Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles PWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D.F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: nolosesamuel@prodigy.net.mx

    2008-07-01

    One of the main challenges in the design of fuel assemblies is the efficient use of uranium achieving burnt homogeneous of the fuel rods as well as the burnt maximum possible of the same ones to the unload. In the case of the assemblies type PWR has been decided actually for fuel assemblies with a single radial enrichment. The present work has like effect to show the because of this decision, reason why a comparison of the neutronic performance of two fuel cells takes place with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The results shown in the present study of the behavior of the neutron flow as well as the power distribution through of assembly sustain the because of a single radial enrichment. (Author)

  12. Evaluation of the presence of a burnable absorber in an assembly 3x3 type PWR; Evaluacion de la presencia de un absorbedor quemable en un ensamble 3x3 tipo PWR

    Martinez F, M. A.; Del Valle G, E.; Alonso V, G. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico)]. e-mail: mike_ipn_esfm@hotmail.com

    2008-07-01

    In the present work the effect is evaluated that causes the presence of a burnable absorber in an adjustment of rods of 3x3 of a fuel assembly type PWR using CASMO-4 code, when comparing the infinite multiplication factor and some average cross sections by means of codes MCNP-4A, CASMO-3 and HELIOS. For this evaluation two cases are evaluated: first consists of an adjustment of rods of 3x3 full completely of fuel and the second consists of a central rod full with a burnable absorber type wet annular burnable absorber (WABA) and the remaining full fuel rods. In both cases the enrichment of the fissile isotopes is varied, for two types of fuel, MOX degree armament and UO{sub 2}. (Author)

  13. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor; Ensaios de lixiviacao em produtos cimentados contendo rejeito simulado de concentrado do evaporador de reator PWR

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes, E-mail: hauczmj@cdtn.b, E-mail: jaalmeida@cdtn.b, E-mail: tellocc@cdtn.b, E-mail: fdc@cdtn.b, E-mail: seless@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-10-26

    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

  14. ASTEC V2.0 reactor applications on French PWR 900 MWe accident sequences and comparison with MAAP4

    Lombard, Virginie; Azarian, Garo; Ducousso, Erik; Gandrille, Pascal, E-mail: pascal.gandrille@areva.com

    2014-06-01

    In the frame of the SARNET Severe Accident Network of Excellence an important task of partners is the assessment of the ASTEC integral code, considered today as the European reference code for evaluation of the source term. A code-to-code comparison between ASTEC V2.0 rev1 and MAAP 4.0.7 code versions has been performed by AREVA NP SAS on a French PWR 900 MWe. Two transients have been analyzed, focussing on in-vessel phenomena: total loss of feedwater (H2 sequence in the French nomenclature) and total loss of onsite and offsite power (H3 sequence). The detailed analysis shows an overall good agreement between both code results on thermal-hydraulics, hydrogen production and core degradation phenomena.

  15. The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

    Kelly, G N

    1983-01-01

    The radiological consequences of degraded core accidents postulated for the Sizewell PWR were assessed in an earlier study and the results published in NRPB-R137. Further analyses have since been made by the Central Electricity Generating Board (CEGB) of degraded core accidents which have led to a revision of their predicted frequencies of occurrence. The implications of these revised frequencies, in terms of the risk to the public from degraded core accidents, are evaluated in this report. Increases, by factors typically within the range of about 1.5 to 7, are predicted in the consequences, compared with those estimated in the earlier study. However, the predicted risk from degraded core accidents, despite these increases, remains exceedingly small.

  16. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    Hartini, Entin; Andiwijayakusuma, Dinan

    2014-09-01

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  17. The development and verification of thermal-hydraulic code on passive residual heat removal system of Chinese advanced PWR

    2006-01-01

    The technology of passive safety is the current trend among safety systems in nuclear power plant. Passive residual heat removal system (PRHRS), a major part of passive safety systems of Chinese advanced PWR, is a novel design with three-fold natural circulation. On the basis of reasonable physics and mathematics models, MITAP-PRHRS code was developed to analyze steady and transient characteristics of the PRHRS. The calculation and analysis show that the code simulates steady characteristics of the PRHRS very well, and it is able to simulate transient characteristics of all startup modes of the PRHRS. However, the quantitative description is poor during the initial stages of the transition process when water hammer occurs.

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

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

    2015-09-15

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

  19. Evaluation of fretting failures on PWR fuel by post-irradiation examinations and modeling in the DEGRAD-1 code

    Castanheira, Myrthes; Silva, Jose Eduardo Rosa da; Lucki, Georgi; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail: myrthes@ipen.br

    2007-07-01

    One of the major recognized causes of fuel rod failures is fretting of the clad due to the entrapment of debris in a fuel rod spacer. Such debris, inadvertently dropped into the primary system during maintenance operations, includes various sizes of particles. Intermediate size particles, such as metal cuttings, electrical connectors, metal fittings, pieces of wire, and small nuts and bolts can become trapped between fuel rods in a spacer where hydraulically induced vibrations can cause fretting failure of the fuel rod. An evaluation of debris fretting failure on PWR fuel is presented. The inquiries on fuel rods failures are based on results of analysis using post-irradiation non-destructive examination. The complementary analysis includes a modeling approach by code DEGRAD-1 to characterize the degradation phenomenon after primary failure integrated in the reactor operational history. (author)

  20. PWR neutron ex-vessel detection calculations using three-dimensional codes; Calculs de detection neutronique externe dans un rep

    Dekens, O.; Lefebvre, J.C.; Rohart, M. [Electricite de France (EDF), 69 -Villeurbanne (France); Chiron, M. [CEA Centre d`Etudes de Saclay, 91 -Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Wouters, R. de [TRACTEBEL, Brussels (Belgium)

    1997-10-01

    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l`Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors). 11 refs.

  1. Practical Application of the MFM Suite on a PWR System: Modelling and Reasoning on Causes and Consequences of Process Anomalies

    Zhang, Xinxin; Thunem, Harald P - J; Lind, Morten;

    2014-01-01

    is equipped with an MFM Model Editing Interface to facilitate the modelling process and MFM model analysis modules to run diag nosis and prognosis analyses based on developed models. New features of the MFM Suite also include making corresponding process diagram for the plant being modelled with MFM......Multilevel Flow Modelling (MFM) is a functional modelling methodology which applies means - end and parts - whole decomposition and aggregation techniques to handle the complexity of engineering systems. It has been adopted in several case studies to model the process goal and functions of PWR...... and linking the MFM model to its process components. The purpose of this report is to make a comprehensive demonstration of how to use the MFM Suite to develop MFM models and run causal reasoning for abnormal situations. This report will explain the capability of representing process and operational knowledge...

  2. Valve inlet fluid conditions for pressurizer safety and relief valves in combustion engineering-designed plants. Final report. [PWR

    Bahr, J.; Chari, D.; Puchir, M.; Weismantel, S.

    1982-12-01

    The purpose of this study is to assemble documented information for C-E designed plants concerning pressurizer safety and power operated relief valve (PROV) inlet fluid conditions during actuation as calculated by conventional licensing analyses. This information is to be used to assist in the justification of the valve inlet fluid conditions selected for the testing of safety valves and PORVs in the EPRI/PWR Safety/Relief Valve Test Program. Available FSAR/Reload analyses and certain low temperature overpressurization analyses were reviewed to identify the pressurization transients which would actuate the valves, and the corresponding valve inlet fluid conditions. In addition, consideration was given to the Extended High Pressure Liquid Injection event. A general description of each pressurization transient is provided. The specific fluid conditions identified and tabulated for each C-E designed plant for each transient are peak pressurizer pressure, pressure ramp rate at actuation, temperature and fluid state.

  3. Development of the new basic correlation “MG-S” for CHF prediction of the PWR fuels

    Yodo, T.; Sato, Y.; Yumura, T.; Makino, Y.; Suemura, T. [Mitsubishi Heavy Industries, LTD., Kobe, Hyogo (Japan)

    2011-07-01

    It is important for core thermal-hydraulic design and plant safety analysis of PWR (Pressurized Water Reactor) to predict CHF (Critical Heat Flux) accurately. The accurate CHF prediction can enhance the reliability of the safety analysis and bring more efficient plant operations such as up-rating and higher burn-up fuel management. The new CHF correlation, MG-S (Mitsubishi Generalized correlation - for Standard grid), has been developed as a basic correlation of the new correlation series, which are for conventional and new-generation Mitsubishi fuel assemblies. Through comparisons with existing CHF data and a conventional CHF correlation, it was confirmed that MG-S can predict CHF with sufficient accuracy and extend its applicability to wider fluid parameters of interest. (author)

  4. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    Hartini, Entin, E-mail: entin@batan.go.id; Andiwijayakusuma, Dinan, E-mail: entin@batan.go.id [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)

    2014-09-30

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  5. Multi level optimization of burnable poison utilization for advanced PWR fuel management

    Yilmaz, Serkan

    The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

  6. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov; Menlove, Howard O., E-mail: hmenlove@lanl.gov

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties.

  7. Application of noble metals on line in Cofrentes NPP and operation experience; Aplicacion de metales nobles en linea en C.N. Cofrentes y experiencia de operacion

    Sanchez Zapata, J. D.

    2015-07-01

    Cofrentes NPP implemented in 2010 the Noble Metal Chemistry as a mitigation technique for the Primary System materials protection against IGSCC. the paper describes briefly the technology fundamentals, the implementation of the specific project, the initial application and the operating experience along the last 3 cycles of the plant. (Author)

  8. Project equipment HVAC chilled with R22 in the NPP Asco; Proyecto de sustitucion de equipos HVAC refrigerados con R22 en la C.N. Asco

    Jaimot Jimenez, J. J.; Imbert, J. J.

    2013-07-01

    This paper describes the project of changing units of air conditioning in the Asco NPP currently used R22 as coolant. The project has a powerful, affecting 37 units, of which more than half are Clase1E. The document describes the process of sizing, scope of change and solutions adopted for this change of design.

  9. A Radiological Survey Approach to Use Prior to Decommissioning: Results from a Technology Scanning and Assessment Project Focused on the Chornobyl NPP

    Milchikov, A.; Hund, G.; Davidko, M.

    1999-10-20

    The primary objectives of this project are to learn how to plan and execute the Technology Scanning and Assessment (TSA) approach by conducting a project and to be able to provide the approach as a capability to the Chernobyl Nuclear Power Plant (ChNPP) and potentially elsewhere. A secondary objective is to learn specifics about decommissioning and in particular about radiological surveying to be performed prior to decommissioning to help ChNPP decision makers. TSA is a multi-faceted capability that monitors and analyzes scientific, technical, regulatory, and business factors and trends for decision makers and company leaders. It is a management tool where information is systematically gathered, analyzed, and used in business planning and decision making. It helps managers by organizing the flow of critical information and provides managers with information they can act upon. The focus of this TSA project is on radiological surveying with the target being ChNPP's Unit 1. This reactor was stopped on November 30, 1996. At this time, Ukraine failed to have a regulatory basis to provide guidelines for nuclear site decommissioning. This situation has not changed as of today. A number of documents have been prepared to become a basis for a combined study of the ChNPP Unit 1 from the engineering and radiological perspectives. The results of such a study are expected to be used when a detailed decommissioning plan is created.

  10. Experimental Studies for the VVER-440/213 Bubble Condenser System for Kola NPP at the Integral Test Facility BC V-213

    Vladimir N. Blinkov

    2012-01-01

    Full Text Available In the frame of Tacis Project R2.01/99, which was running from 2003 to 2005, the bubble condenser system of Kola NPP (unit 3 was qualified at the integral test facility BC V-213. Three LB LOCA tests, two MSLB tests, and one SB LOCA test were performed. The appropriate test scenarios for BC V-213 test facility, modeling accidents in the Kola NPP unit 3, were determined with pretest calculations. Analysis of test results has shown that calculated initial conditions and test scenarios were properly reproduced in the tests. The detailed posttest analysis of the tests performed at BC V-213 test facility was aimed to validate the COCOSYS code for the calculation of thermohydraulic processes in the hermetic compartments and bubble condenser. After that the validated COCOSYS code was applied to NPP calculations for Kola NPP (unit 3. Results of Tacis R2.01/99 Project confirmed the bubble condenser functionality during large and small break LOCAs and MSLB accidents. Maximum loads were reached in the LB LOCA case. No condensation oscillations were observed.

  11. Cytogenetic examination of persons working in the area of radiation accident at the Fukushima-1 NPP in Japan

    Nugis V.Yu.

    2014-12-01

    Full Text Available Purpose: biological dose indication of employees of the Ministry of Emergency Situations of Russia who took part in the work in Japan in connection with the accident at Fukushima-1 NPP and several journalists covering this event. Material and methods. The analysis of chromosomal aberrations in peripheral blood lymphocyte cultures of 46 people was performed. Results. The frequency of chromosomal damages exceeded background levels in only 3 people, and aberration character testified irradiation in previous situations. Conclusion. The significant overexposure of these workers during they stayed in Japan is absent, however it is necessary to perform a preliminary analysis of chromosome aberrations if you intend to exercise of biological dose indication after returning of people from areas of potential exposure.

  12. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Kral, P. [Nuclear Research Inst. Rez (Switzerland)

    1995-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  13. Sensitivity analysis of the spectra of the core neutronic source in the calculation of radiation damage in internal of PWR reactor vessel. Internal; Analisis de sensibilidad a los espectros de la fuente neutronica del nucleo en el calculo del dano por irradiacion en los internos de la vasija de un reactor PWR

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barrerira Pereira, P.

    2012-07-01

    This study is to analyze the sensitivity to the expected differences in the energy spectra characterizing the neutron source that radiates the vessel internals of a commercial PWR reactor, in order to quantify their influence in the quantities that determine the damage in materials metal.

  14. Thermal-hydraulic analysis best-estimate of an accident in the containment a PWR-W reactor with GOTHIC code using a 3D model detailed; Analisis termo-hidraulico best-estimate de un accidente en contencion de un reactor PWR-W con el codigo GOTHIC mediante un modelo 3D detallado

    Bocanegra, R.; Jimenez, G.

    2013-07-01

    The objective of this project will be a model of containment PWR-W with the GOTHIC code that allows analyzing the behavior detailed after a design basis accident or a severe accident. Unlike the models normally used in codes of this type, the analysis will take place using a three-dimensional model of the containment, being this much more accurate.

  15. Radiometric Inter-Calibration between Himawari-8 AHI and S-NPP VIIRS for the Solar Reflective Bands

    Fangfang Yu

    2016-02-01

    Full Text Available The Advanced Himawari Imager (AHI on-board Himawari-8, which was launched on 7 October 2014, is the first geostationary instrument housed with a solar diffuser to provide accurate onboard calibrated data for the visible and near-infrared (VNIR bands. In this study, the Ray-matching and collocated Deep Convective Cloud (DCC methods, both of which are based on incidently collocated homogeneous pairs between AHI and Suomi NPP (S-NPP Visible Infrared Imaging Radiometer Suite (VIIRS, are used to evaluate the calibration difference between these two instruments. While the Ray-matching method is used to examine the reflectance difference over the all-sky collocations with similar viewing and illumination geometries, the near lambertian collocated DCC pxiels are used to examine the difference for the median or high reflectance scenes. Strong linear relationships between AHI and VIIRS can be found at all the paired AHI and VIIRS bands. Results of both methods indicate that AHI radiometric calibration accuracy agrees well with VIIRS data within 5% for B1-4 and B6 at mid and high reflectance scenes, while AHI B5 is generally brighter than VIIRS by ~6%–8%. No apparent East-West viewing angle dependent calibration difference can be found at all the VNIR bands. Compared to the Ray-matching method, the collocated DCC method provides less uncertainty of inter-calibration results at near-infrared (NIR bands. As AHI has similar optics and calibration designs to the GOES-R Advanced Baseline Imager (ABI, which is currently scheduled to launch in fall 2016, the on-orbit AHI data provides a unique opportunity to develop, test and examine the cal/val tools developed for ABI.

  16. Inter-comparison of NPP/CrIS radiances with VIIRS, AIRS, and IASI: a post-launch calibration assessment

    Wang, Likun; Han, Yong; Tremblay, Denis; Weng, Fuzhong; Goldberg, Mitchell

    2012-11-01

    The Cross-track Infrared Sounder (CrIS) on the newly-launched Suomi National Polar-orbiting Partnership (Suomi NPP) is a Fourier transform spectrometer that provides soundings of the atmosphere with 1305 spectral channels, over 3 wavelength ranges: LWIR (9.14 - 15.38 μm) MWIR (5.71 - 8.26 μm) and SWIR (3.92 - 4.64 μm). An accurate spectral and radiometric calibration as well as geolocation is fundamental for CrIS radiance Sensor Data Records (SDRs). In this study, through inter- and intra-satellite calibration efforts, we focus on assessment of NPP/CrIS post-launch performance. First, we compare CrIS hyperspectral radiance measurements with the Atmospheric Infrared Sounder (AIRS) on NASA Earth Observing System (EOS) Aqua and Infrared Atmospheric Sounding Interferometer (IASI) on Metop-A to examine spectral and radiometric consistence and difference among three hyperspectral IR sounders. Secondly, an accurate collocation algorithm has been developed to collocate high spatial resolution measurements from the Visible Infrared Imager Radiometer Suite (VIIRS) within each CrIS Field of View (FOV). We compare CrIS spectrally-averaged radiances with the spatially-averaged and collocated pixels from the VIIRS IR channels. Since CrIS and VIIRS are onboard on the same satellite platform, the intra-satellite comparison will allow examining the radiometric difference between CrIS and VIIRS with scene temperatures, scan angles, and orbital position. In addition, given a high spatial resolution of VIIRS channels, the VIIRS-CrIS comparison results can access geolocation accuracy of CrIS that have relatively large FOVs (14 km at ndair) using high resolution VIIRS pixel (375m or 750m at nadir).

  17. Towards remote sensing of Arctic ice roads and associated human activities using SUOMI NPP night light images

    Bennett, M.; Smith, L. C.; Stephenson, S. R.

    2014-12-01

    Ice roads are often the only cost-effective means of transporting goods and supplies to communities, mines, and other sites in remote parts of the Arctic. Yet, there is no global dataset for Arctic ice roads. However, remotely sensed images from the SUOMI NPP day/night band (DNB) of the Visible Infrared Imaging Radiometer Suite (VIIRS) may allow for the construction of such a dataset. The DNB's high sensitivity to low-level light suggests that while it is not feasible to view ice roads at night per se, other prominent features associated with ice roads can serve as proxies. Using a time series of images taken in winter 2012, 2013, and 2014, SUOMI NPP images are compared with Landsat 8 images and an existing map of the Tibbitt to Contwoyto ice road in the Northwest Territories and Nunavut, Canada. First results reveal that while the ice road's exact path cannot be discerned, key points of human activity along the way can be made out. This bodes well for future applications of DNB imagery to detect ice roads in places like the Russian Federation, for which there is a dearth of publicly available maps. Knowing the location of ice roads is important for two reasons. First, these data can signal sites of natural resource extraction in places for which information is not widely disseminated, such as in the Russian Far East. Second, new geospatial datasets for ice roads can be combined with models assessing impacts of climate change on circumpolar land accessibility (Stephenson et al. 2011) in order to understand where the structural integrity of ice roads may be at risk. As warming temperatures threaten to shorten the season for ice roads, communities and mines alike will need to prepare for changes to their transportation infrastructure, made out of the changing landscape itself.

  18. Assessment of subcriticality during PWR-type reactor refueling; Evaluation de la sous-criticite lors des operations de chargement d'un reacteur nucleaire REP

    Verdier, A

    2005-04-15

    During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-{alpha} and Feynman-{alpha} methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

  19. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  20. Technical basis for the initiation and cessation of environmentally-assisted cracking of low-alloy steels in elevated temperature PWR environments

    James, L.A.

    1997-10-01

    The Section 11 Working Group on Flaw Evaluation of the ASME B and PV Code Committee is considering a Code Case to allow the determination of the conditions under which environmentally-assisted cracking of low-alloy steels could occur in PWR primary environments. This paper provides the technical support basis for such an EAC Initiation and Cessation Criterion by reviewing the theoretical and experimental information in support of the proposed Code Case.

  1. The distribution of radionuclides between the sediments and macrophytes in the cooling pond of the Ignalina NPP - The Distribution of Radionuclides in Freshwater Hydro Ecosystem's Bottom Sediments and Macrophytes depending on the Ecological Conditions

    Marciulioniene, D.; Jefanova, O.; Mazeika, J. [Nature Research Centre, Akademijos str. 2, LT-08412 Vilnius, Lietuva (Lithuania)

    2014-07-01

    The distribution of {sup 137}Cs, {sup 60}Co, {sup 54}Mn in the aquatory of lake Drukshiai (the monitoring stations), the coastal area of this lake, the industrial drainage systems channel of the Ignalina NPP and the cooling water channel of the Ignalina NPP was analyzed on the basis of long-term (1988-2009) investigations of radionuclides specific activity in bottom sediments and macrophytes, also the ability of radionuclides falling into lake Drukshiai from the Ignalina NPP through effluents channels was assessed. It was established that {sup 137}Cs, {sup 60}Co and {sup 54}Mn in the bottom sediments and the macrophytes were distributed quite differently in the monitoring stations of lake Drukshiai and the coastal area as well as in the industrial drainage systems channel of the Ignalina NPP and the cooling water channel of the Ignalina NPP. The different characteristics of the sediments, various ecological conditions, as well as the existing anthropogenic environmental factors and the different in the ecological groups of the plants could have had impact on the distribution of {sup 137}Cs, {sup 60}Co and {sup 54}Mn in the bottom sediments and the aquatic plants in lake Drukshiai and the effluents channels of the Ignalina NPP. The {sup 137}Cs, {sup 60}Co and {sup 54}Mn specific activity's values were significantly higher in macrophytes from the industrial drainage systems channel of Ignalina NPP than in macrophytes from the cooling water channel. Nevertheless the specific activities level of these radionuclides differed only slightly in the macrophytes from the areas which were impacted by the effluents channels of the Ignalina NPP. This can be explained by the fact that the phyto-remediation (as the form of auto-purification) of these effluents from the radionuclides had been present in the industrial drainage systems channel of Ignalina NPP before entering the water into lake Drukshiai. (authors)

  2. Acceptance test for 900 MWe PWR unit replacement steam generators; Essai de reception des generateurs de vapeur de remplacement des tranches REP 900

    Gourguechon, B.

    1993-12-31

    During the first half of 1994, the Gravelines 1 steam generators will be replaced (SG replacement procedure). The new SG`s differ from the former components notably by the alloy used for the tube bundle, in this case, the high chromium content Inconel 690. So, from this standpoint, they are to be considered as PWR 900 replacement SG first models and their thermal efficiency has consequently to be assessed. This will provide an opportunity of ensuring that the performance of the components delivered is in compliance with requirements and of making the necessary provisions if significant deviations are observed. The EFMT branch, which has been in charge of the instrumentation and acceptance of the different SG first models since the first PWR plants were commissioned, will be responsible for the acceptance tests and the ultimate validation of a performance assessment procedure applicable to the future replacement steam generators. The methods and tests proposed for SG expert appraisal are based on consideration of the importance of primary measurement quality for satisfactory SG assessment and of the new test facilities with which the 900 and 1 300 PWR plants are gradually being equipped. These facilities provide an on-site computer environment for tests compatible with the tools (PATTERN, etc.) used at EFMT and in other departments. This test is the first of this kind performed by EFMT and the test facility of a nuclear power plant. (author). 6 figs.

  3. Oxide property of SG tube materials exposed to an alkaline environment as a secondary side of a PWR

    Kim, Dongjin; Mun, Byung Hak; Kim, Hong Pyo; Hwang, Seong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Stress corrosion cracking (SCC) is an issue that should be overcome in nuclear power plants (NPP). Recognizing that cracks initiate and propagate through unavoidable breakdowns and alterations of the surface oxide on Alloy 600, the SCC behavior is closely related to the oxide property. Corrosion resistance against SCC, in particular, was improved through a newly developed heat treatment process from LTMA (low temperature mill annealed) Alloy 600 to HTMA (high temperature mill annealed) Alloy 600, and then TT (thermally treated) Alloy 600. Intra-granular carbide widely spread in LTMA Alloy 600 dissolves, and inter-granular carbide is then formed during high-temperature mill annealing and cooling, which leads to a great SCC resistance enhancement. Inter-granular carbide is well developed, healing chromium depletion at a grain boundary, and residual stress is removed during additional thermal treatment following mill annealing, which improves the SCC resistance more. In spite of this improvement of TT Alloy 600, Seabrook and Vogtle 1 in the US, using TT Alloy 600, also showed SCC due to a non-optimum microstructure, residual stress, Pb existence, and so on over a 20-year operation of an NPP even though SCC occurs less frequently than LTMA and (or) HTMA Alloy 600s. SCC has also occurred for TT Alloy 600 tubes in Korea, whose main causes resemble US cases. The pH at high temperature in the crevice of SG tubes distributes from acidic of 4 to alkaline above 10 at high temperature depending on the impurity concentration such as chloride and hydroxide ions including other corrosive impurities such as Pb known as very detrimental species even though the bulk pH of secondary water is a mild alkaline solution. Regarding the aggressiveness of Pb, even Alloy 690 is also susceptible to SCC in a strong alkaline solution with lead. Therefore, in the present work, the oxides were investigated in a leaded alkaline solution of pH(T) 9.9 at 315 .deg. C as a function of immersion time

  4. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  5. Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw; Hsu, Keng-Hsien, E-mail: hardlycampus@iner.gov.tw; Lin, Chin-Tsu, E-mail: jtling@iner.gov.tw

    2015-07-15

    Highlights: • Calculate NSSS and containment transient response during extended SBO of 24 h. • RELAP5-3D and GOTHIC models are developed for Maanshan PWR plant. • Reactor coolant pump seal leakage is specifically modeled for each loop. • Analyses are performed with and without secondary-side depressurization, respectively. • Considering different total available time for turbine driven auxiliary feedwater system. - Abstract: A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP5-3D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS

  6. High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit

    2014-08-01

    In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover

  7. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

    Monteiro, Iara Arraes

    1999-02-15

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  8. GHRSST 2 Level 2P Global Skin Sea Surface Temperature from the Visible Infrared Imaging Radiometer Suite (VIIRS) on the Suomi NPP satellite created by the NOAA Advanced Clear-Sky Processor for Ocean (ACSPO) (GDS version 2)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Joint Polar Satellite System (JPSS), starting with S-NPP launched on 28 October 2011, is the new generation of the US Polar Operational Environmental Satellites...

  9. VERONA V6.22 – An enhanced reactor analysis tool applied for continuous core parameter monitoring at Paks NPP

    Végh, J., E-mail: janos.vegh@ec.europa.eu [Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Pós, I., E-mail: pos@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Horváth, Cs., E-mail: csaba.horvath@energia.mta.hu [Centre for Energy Research, Hungarian Academy of Sciences, H-1525 Budapest 114, P.O. Box 49 (Hungary); Kálya, Z., E-mail: kalyaz@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Parkó, T., E-mail: parkot@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Ignits, M., E-mail: ignits@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary)

    2015-10-15

    Between 2003 and 2007 the Hungarian Paks NPP performed a large modernization project to upgrade its VERONA core monitoring system. The modernization work resulted in a state-of-the-art system that was able to support the reactor thermal power increase to 108% by more accurate and more frequent core analysis. Details of the new system are given in Végh et al. (2008), the most important improvements were as follows: complete replacement of the hardware and the local area network; application of a new operating system and porting a large fraction of the original application software to the new environment; implementation of a new human-system interface; and last but not least, introduction of new reactor physics calculations. Basic novelty of the modernized core analysis was the introduction of an on-line core-follow module based on the standard Paks NPP core design code HELIOS/C-PORCA. New calculations also provided much finer spatial resolution, both in terms of axial node numbers and within the fuel assemblies. The new system was able to calculate the fuel applied during the first phase of power increase accurately, but it was not tailored to determine the effects of burnable absorbers as gadolinium. However, in the second phase of the power increase process the application of fuel assemblies containing three fuel rods with gadolinium content was intended (in order to optimize fuel economy), therefore off-line and on-line VERONA reactor physics models had to be further modified, to be able to handle the new fuel according to the accuracy requirements. In the present paper first a brief overview of the system version (V6.0) commissioned after the first modernization step is outlined; then details of the modified off-line and on-line reactor physics calculations are described. Validation results for new modules are treated extensively, in order to illustrate the extent and complexity of the V&V procedure associated with the development and licensing of the new

  10. Radiation conditions in the Oryol region territory impacted by radioactive contamination caused by the Chernobyl NPP accident

    G. L. Zakharchenko

    2016-01-01

    Full Text Available Research objective is retrospective analysis of radiation conditions in the Oryol region during 1986- 2015 and assessment of efficacy of the carried out sanitary and preventive activities for population protection against radiation contamination caused by the Chernobyl NPP accident.Article materials were own memoirs of events participants, analysis of federal state statistic surveillance forms 3-DOZ across the Oryol region, f-35 “Data on patients with malignant neoplasms, f-12 “Report on MPI activities”. Risk assessment of oncological diseases occurrence is carried out on the basis of AAED for 1986- 2014 using the method of population exposure risk assessment due to long uniform man-made irradiation in small doses. Results of medical and sociological research of genetic, environmental, professional and lifestyle factors were obtained using the method of cancer patients’ anonymous survey. Data on "risk" factors were obtained from 467 patients hospitalized at the Budgetary Health Care Institution of the Oryol region “Oryol oncology clinic”; a specially developed questionnaire with 60 questions was filled out.The article employs the method of retrospective analysis of laboratory and tool research and calculation of dose loads on the Oryol region population, executed throughout the whole period after the accident.This article provides results of the carried out laboratory research of foodstuff, environment objects describing the radiation conditions in the Oryol region since the first days after the Chernobyl NPP accident in 1986 till 2015.We presented a number of activities aimed at liquidation of man-caused radiation accident consequences which were developed and executed by the experts of the Oryol region sanitary and epidemiology service in 1986-2015. On the basis of the above-stated one may draw the conclusions listed below. Due to interdepartmental interaction and active work of executive authorities in the Oryol region, the

  11. Consistent High-Quality Global SO2 and HCHO Datasets from EOS Aura/OMI and Suomi NPP/OMPS

    Li, C.; Joiner, J.; Krotkov, N. A.; Fioletov, V.; McLinden, C. A.; Zhang, Y.

    2015-12-01

    We report on recent effort and progress at NASA Goddard Space Flight Center in developing consistent SO2 and HCHO retrieval products from Aura/Ozone Monitoring Instrument (OMI) and Suomi National Polar-orbiting Partnership (S-NPP)/Ozone Mapping and Profiler Suite (OMPS) nadir mapper. Given the substantial differences between OMI and OMPS in several key aspects, such as spatial and spectral resolution and signal-to-noise ratio, a major challenge in ensuring data continuity between the two instruments is to properly account for different instrument characteristics as well as instruments' degradation over time. To this end, we have developed an innovative approach based on principal component analysis (PCA) of measured Earthshine radiances. We utilize a PCA technique to extract a series of spectral features (principal components or PCs) explaining the variance of measured reflectance spectra, associated with both physical processes (e.g., ozone absorption, rotational Raman scattering) and measurement details (e.g., wavelength shift). By fitting these PCs along with pre-computed Jacobians for our target species (SO2 or HCHO) to the measured radiance spectra, we can estimate the atmospheric loading of SO2 or HCHO while minimizing the impacts of interfering processes and measurement imperfection on retrievals. Since no explicit instrument-specific radiance data correction scheme is required, the PCA method is easily implemented with both OMI and OMPS and maximizes data continuity. The PCA algorithm currently runs operationally in the production of the new generation NASA standard OMI planetary boundary layer (PBL) SO2 data that have been shown to improve the detection limit of anthropogenic SO2 emission sources by a factor of two, as compared with the previous generation product. In this presentation, we will demonstrate that the PCA algorithm can produce SO2 and HCHO retrievals from OMPS that have comparable data quality with our OMI retrievals. We will also demonstrate

  12. Decommissioning of the NPP Obrigheim (KWO). Shutdown/close-down of systems or components; Stilllegungsbetrieb der Anlage KWO Obrigheim. Ausserbetriebnahme / Stillsetzung von Systemen oder Anlagenteilen

    Rausch, Eberhard H. [ISE Ingenieurgesellschaft fuer Stilllegung und Entsorgung mbH, Roedermark (Germany); Rudolf, Dieter [Energie Baden-Wuerttemberg AG (EnBW), Karlsruhe (Germany)

    2012-11-01

    As a consequence of the decommissioning of the NPP Obrigheim (KWO) the plant was transferred into the decommissioning operation, including the operation of several safety relevant systems and the storage of irradiation fuel elements. Actually, the fuel element have been removed from the reactor pressure vessel and the reactor building 01 and are now stored in an external fuel element storage pool at the NPP site. Most of the systems required for power operation have been shutdown (drained, depressurized, cold, and disconnected from operated systems). The operated systems exhibit significantly lower working pressure and temperatures compared to power operation. The shutdown is performed stepwise, for each system a shutdown plan has to be prepared, describing the scheduled measures. The presentation includes details of the work flow of the performed and planned system shutdown.

  13. Effect of water chemistry on environmentally assisted cracking of alloy 600 in simulated primary side PWR environments

    Koenig, M. [Studsvik Nuvlear (Sweden); Lidar, P. [GSE Power Systems (Sweden); Engstroem, J. [Ringhals NPP (Sweden); Gott, K. [SKI Sweden (Sweden)

    2002-07-01

    Environmental aspects of crack growth due to intergranular stress corrosion cracking (IGSCC) of Alloy 600 in simulated primary side PWR environments have been studied. The purpose of the study was to quantify the effects of the water chemistry (Li, B and H{sub 2} concentrations, and the pH-value by adding KOH) on the crack growth rate, da/dt. 12.5 mm thick compact tension (CT) specimens were used for testing at a constant maximum stress intensity factor in the range of 26-32 MPa{open_square}m. The crack growth was continuously monitored using a direct current potential drop system. Intergranular crack growth due to IGSCC was dominant in the specimens, although there were also small fractions of transgranular cracking. Multivariate analysis was used on the results from the present work together with results from previous tests on the same material. Temperature and the stress intensity were also included as factors in the analysis. A partial least squares regression was developed and interaction effects between the factors were found to affect the crack growth rate. The Partial Least Square regression predicts the observed crack growth rates reasonably well. (authors)

  14. The prediction of pH by Gibbs free energy minimization in the sump solution under LOCA condition of PWR

    Yoon, Hyoung Ju [Dept. of Nuclear Engineering, University of Kyunghee, Seoul (Korea, Republic of)

    2013-02-15

    It is required that the pH of the sump solution should be above 7.0 to retain iodine in a liquid phase and be within the material compatibility constraints under LOCA condition of PWR. The pH of the sump solution can be determined by conventional chemical equilibrium constants or by the minimization of Gibbs free energy. The latter method developed as a computer code called SOLGASMIX-PV is more convenient than the former since various chemical components can be easily treated under LOCA conditions. In this study, SOLGASMIX-PV code was modified to accommodate the acidic and basic materials produced by radiolysis reactions and to calculate the pH of the sump solution. When the computed pH was compared with measured by the ORNL experiment to verify the reliability of the modified code, the error between two values was within 0.3 pH. Finally, two cases of calculation were performed for the SKN 3 and 4 and UCN 1 and 2. As results, pH of the sump solution for the SKN 3 and 4 was between 7.02 and 7.45, and for the UCN 1 and 2 plant between 8.07 and 9.41. Furthermore, it was found that the radiolysis reactions have insignificant effects on pH because the relative concentrations of HCl, HNO3, and Cs are very low.

  15. A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

    Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

  16. Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel

    Hermann, O.W.

    2000-02-01

    The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

  17. Containment Depressurization Capabilities of Filtered Venting System in 1000 MWe PWR with Large Dry Containment

    Sang-Won Lee

    2014-01-01

    Full Text Available After the Fukushima Daiichi nuclear power plant accident, the Korean government and nuclear industries performed comprehensive safety inspections on all domestic nuclear power plants against beyond design bases events. As a result, a total of 50 recommendations were defined as safety improvement action items. One of them is installation of a containment filtered venting system (CFVS or portable backup containment spray system. In this paper, the applicability of CFVS is examined for OPR1000, a 1000 MWe PWR with large dry containment in Korea. Thermohydraulic analysis results show that a filtered discharge flow rate of 15 [kg/s] at 0.9 [MPa] is sufficient to depressurize the containment against representative containment overpressurization scenarios. Radiological release to the environment is reduced to 10-3 considering the decontamination factor. Also, this cyclic venting strategy reduces noble gas release by 50% for 7 days. The probability of maintaining the containment integrity in level 2 probabilistic safety assessment (PSA initiating events is improved twofold, from 43% to 87%. So, the CFVS can further improve the containment integrity in severe accident conditions.

  18. TAPINS: A THERMAL-HYDRAULIC SYSTEM CODE FOR TRANSIENT ANALYSIS OF A FULLY-PASSIVE INTEGRAL PWR

    YEON-GUN LEE

    2013-08-01

    Full Text Available REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility. Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

  19. THE PREDICTION OF pH BY GIBBS FREE ENERGY MINIMIZATION IN THE SUMP SOLUTION UNDER LOCA CONDITION OF PWR

    HYOUNGJU YOON

    2013-02-01

    Full Text Available It is required that the pH of the sump solution should be above 7.0 to retain iodine in a liquid phase and be within the material compatibility constraints under LOCA condition of PWR. The pH of the sump solution can be determined by conventional chemical equilibrium constants or by the minimization of Gibbs free energy. The latter method developed as a computer code called SOLGASMIX-PV is more convenient than the former since various chemical components can be easily treated under LOCA conditions. In this study, SOLGASMIX-PV code was modified to accommodate the acidic and basic materials produced by radiolysis reactions and to calculate the pH of the sump solution. When the computed pH was compared with measured by the ORNL experiment to verify the reliability of the modified code, the error between two values was within 0.3 pH. Finally, two cases of calculation were performed for the SKN 3&4 and UCN 1&2. As results, pH of the sump solution for the SKN 3&4 was between 7.02 and 7.45, and for the UCN 1&2 plant between 8.07 and 9.41. Furthermore, it was found that the radiolysis reactions have insignificant effects on pH because the relative concentrations of HCl, HNO3, and Cs are very low.

  20. Presentation of the MERC work-flow for the computation of a 2D radial reflector in a PWR

    Clerc, T.; Hebert, A. [Institut de Genie Nucleaire, Station Centre-Ville, Montreal, QC, H3C 3A7 (Canada); Leroyer, H.; Argaud, J. P.; Poncot, A.; Bouriquet, B. [Electricite de France, R and D, SINETICS, 1 Av. du General de Gaulle, 92141, Clamart (France)

    2013-07-01

    This paper presents a work-flow for computing an equivalent 2D radial reflector in a pressurized water reactor (PWR) core, in adequacy with a reference power distribution, computed with the method of characteristics (MOC) of the lattice code APOLLO2. The Multi-modelling Equivalent Reflector Computation (MERC) work-flow is a coherent association of the lattice code APOLLO2 and the core code COCAGNE, structured around the ADAO (Assimilation de Donnees et Aide a l'Optimisation) module of the SALOME platform, based on the data assimilation theory. This study leads to the computation of equivalent few-groups reflectors, that can be spatially heterogeneous, which have been compared to those obtained with the OPTEX similar methodology developed with the core code DONJON, as a first validation step. Subsequently, the MERC work-flow is used to compute the most accurate reflector in consistency with all the R and D choices made at Electricite de France (EDF) for the core modelling, in terms of number of energy groups and simplified transport solvers. We observe important reductions of the power discrepancies distribution over the core when using equivalent reflectors obtained with the MERC work-flow. (authors)

  1. Large Scale Finite Element Thermal Analysis of the Bolts of a French PWR Core Internal Baffle Structure

    Rupp, Isabelle; Christophe, Peniguel [EDF R and D, Paris (France); Tommy, Martin Michel [1 av du General de Gaulle, Paris (France)

    2009-11-15

    The internal core baffle structure of a French Pressurized Water Reactor (PWR) consists of a collection of baffles and formers that are attached to the barrel. The connections are done thanks to a large number of bolts (about 1500). After inspection, some of the bolts have been found cracked. This has been attributed to the Irradiation Assisted Stress Corrosion Cracking (IASCC). The Electricite De France (EDF) has set up a research program to gain better knowledge of the temperature distribution, which may affect the bolts and the whole structure. The temperature distribution in the structure was calculated thanks to the thermal code SYRTHES that used a finite element approach. The heat transfer between the by-pass flow inside the cavities of the core baffle and the structure was accounted for thanks to a strong thermal coupling between the thermal code SYRTHES and the CFD code named Code{sub S}aturne. The results for the CP0 plant design show that both the high temperature and strong temperature gradients could potentially induce mechanical stresses. The CPY design, where each bolt is individually cooled, had led to a reduction of temperatures inside the structures. A new parallel version of SYRTHES, for calculations on very large meshes and based on MPI, has been developed. A demonstration test on the complete structure that has led to about 1.1 billion linear tetraedra has been calculated on 2048 processors of the EDF Blue Gene computer

  2. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the STVU content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The STYU content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared.

  3. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

  4. OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle

    A. Del Nevo

    2012-01-01

    Full Text Available Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualification of the traditional methods (system codes can be further enhanced by validation against high-quality experimental data (e.g., including measurement of local parameters. One of these databases, related to the void fraction measurements, is the pressurized water reactor subchannel and bundle tests (PSBT conducted by the Nuclear Power Engineering Corporation (NUPEC in Japan. Selected experiments belonging to this database are used for the OECD/NRC PSBT benchmark. The activity presented in the paper is connected with the improvement of current approaches by comparing system code predictions with measured data on void production in PWR-type fuel bundles. It is aimed at contributing to the validation of the numerical models of CATHARE 2 code, particularly for the prediction of void fraction distribution both at subchannel and bundle scale, for different test bundle configurations and thermal-hydraulic conditions, both in steady-state and transient conditions.

  5. Computational simulation of natural convection of a molten core in lower head of a PWR pressure vessel

    Vieira, Camila Braga; Romero, Gabriel Alves; Jian Su, E-mail: camila@lasme.coppe.ufrj.b, E-mail: gabrielromero@lasme.coppe.ufrj.b, E-mail: sujian@lasme.coppe.ufrj.b [Universidade Federal do Rio de Janeiro (COPPE/UFRJ), RJ (Brazil). Nuclear Engineering Program

    2010-07-01

    Computational simulation of natural convection in a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel was performed for two-dimensional semi-circular geometry with isothermal walls. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate was simulated by using a commercial computational fluid dynamics software ANSYS CFX 12.0. The Boussinesq model was used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-{omega} based SST (Shear Stress Transport) turbulence model was used to mould the turbulent stresses in the Reynolds-Average Navier-Stokes equations (RANS). Two Prandtl numbers, 6:13 and 7:0, were considered. Five Rayleigh numbers were simulated for each Prandtl number used (109, 1010, 1011, 1012, and 1013). The average Nusselt numbers on the bottom surface of the semicircular cavity were in excellent agreement with Mayinger et al. (1976) correlation and only at Ra = 109 the average Nusselt number on the top flat surface was in agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations. (author)

  6. Advanced methods for the study of PWR cores; Les methodes d'etudes avancees pour les coeurs de REP

    Lambert, M.; Salvatores, St.; Ferrier, A. [Electricite de France (EDF), Service Etudes et Projets Thermiques et Nucleaires, 92 - Courbevoie (France); Pelet, J.; Nicaise, N.; Pouliquen, J.Y.; Foret, F. [FRAMATOME ANP, 92 - Paris La Defence (France); Chauliac, C. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), 91 - Gif sur Yvette (France); Johner, J. [CEA Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Cohen, Ch

    2003-07-01

    This document gathers the transparencies presented at the 6. technical session of the French nuclear energy society (SFEN) in October 2003. The transparencies of the annual meeting are presented in the introductive part: 1 - status of the French nuclear park: nuclear energy results, management of an exceptional climatic situation: the heat wave of summer 2003 and the power generation (J.C. Barral); 2 - status of the research on controlled thermonuclear fusion (J. Johner). Then follows the technical session about the advanced methods for the study of PWR reactor cores: 1 - the evolution approach of study methodologies (M. Lambert, J. Pelet); 2 - the point of view of the nuclear safety authority (D. Brenot); 3 - the improved decoupled methodology for the steam pipe rupture (S. Salvatores, J.Y. Pouliquen); 4 - the MIR method for the pellet-clad interaction (renovated IPG methodology) (E. Baud, C. Royere); 5 - the improved fuel management (IFM) studies for Koeberg (C. Cohen); 6 - principle of the methods of accident study implemented for the European pressurized reactor (EPR) (F. Foret, A. Ferrier); 7 - accident studies with the EPR, steam pipe rupture (N. Nicaise, S. Salvatores); 8 - the co-development platform, a new generation of software tools for the new methodologies (C. Chauliac). (J.S.)

  7. The Influences of Drought and Land-Cover Conversion on Inter-Annual Variation of NPP in the Three-North Shelterbelt Program Zone of China Based on MODIS Data.

    Dailiang Peng

    Full Text Available Terrestrial ecosystems greatly contribute to carbon (C emission reduction targets through photosynthetic C uptake.Net primary production (NPP represents the amount of atmospheric C fixed by plants and accumulated as biomass. The Three-North Shelterbelt Program (TNSP zone accounts for more than 40% of China's landmass. This zone has been the scene of several large-scale ecological restoration efforts since the late 1990s, and has witnessed significant changes in climate and human activities.Assessing the relative roles of different causal factors on NPP variability in TNSP zone is very important for establishing reasonable local policies to realize the emission reduction targets for central government. In this study, we examined the relative roles of drought and land cover conversion(LCC on inter-annual changes of TNSP zone for 2001-2010. We applied integrated correlation and decomposition analyses to a Standardized Evapotranspiration Index (SPEI and MODIS land cover dataset. Our results show that the 10-year average NPP within this region was about 420 Tg C. We found that about 60% of total annual NPP over the study area was significantly correlated with SPEI (p<0.05. The LCC-NPP relationship, which is especially evident for forests in the south-central area, indicates that ecological programs have a positive impact on C sequestration in the TNSP zone. Decomposition analysis generally indicated that the contributions of LCC, drought, and other Natural or Anthropogenic activities (ONA to changes in NPP generally had a consistent distribution pattern for consecutive years. Drought and ONA contributed about 74% and 23% to the total changes in NPP, respectively, and the remaining 3% was attributed to LCC. Our results highlight the importance of rainfall supply on NPP variability in the TNSP zone.

  8. The Influences of Drought and Land-Cover Conversion on Inter-Annual Variation of NPP in the Three-North Shelterbelt Program Zone of China Based on MODIS Data.

    Peng, Dailiang; Wu, Chaoyang; Zhang, Bing; Huete, Alfredo; Zhang, Xiaoyang; Sun, Rui; Lei, Liping; Huang, Wenjing; Liu, Liangyun; Liu, Xinjie; Li, Jun; Luo, Shezhou; Fang, Bin

    2016-01-01

    Terrestrial ecosystems greatly contribute to carbon (C) emission reduction targets through photosynthetic C uptake.Net primary production (NPP) represents the amount of atmospheric C fixed by plants and accumulated as biomass. The Three-North Shelterbelt Program (TNSP) zone accounts for more than 40% of China's landmass. This zone has been the scene of several large-scale ecological restoration efforts since the late 1990s, and has witnessed significant changes in climate and human activities.Assessing the relative roles of different causal factors on NPP variability in TNSP zone is very important for establishing reasonable local policies to realize the emission reduction targets for central government. In this study, we examined the relative roles of drought and land cover conversion(LCC) on inter-annual changes of TNSP zone for 2001-2010. We applied integrated correlation and decomposition analyses to a Standardized Evapotranspiration Index (SPEI) and MODIS land cover dataset. Our results show that the 10-year average NPP within this region was about 420 Tg C. We found that about 60% of total annual NPP over the study area was significantly correlated with SPEI (p<0.05). The LCC-NPP relationship, which is especially evident for forests in the south-central area, indicates that ecological programs have a positive impact on C sequestration in the TNSP zone. Decomposition analysis generally indicated that the contributions of LCC, drought, and other Natural or Anthropogenic activities (ONA) to changes in NPP generally had a consistent distribution pattern for consecutive years. Drought and ONA contributed about 74% and 23% to the total changes in NPP, respectively, and the remaining 3% was attributed to LCC. Our results highlight the importance of rainfall supply on NPP variability in the TNSP zone.

  9. Technical aspects of the process of segmentation and packaging of the reactor vessel of Jose Cabrera NPP; Aspectos tecnicos del proceso de segmentacion y embalaje de la vasija del reactor de la central nuclear Jose Cabrera

    Valdivieso, J. M.; Garcia Castro, R.

    2015-07-01

    Westinghouse is carrying out the segmentation of the reactor pressure vessel (RPV) within the framework of the Dismantling and Decommissioning Project of the Jose Cabrera NPP. The final concept is based on the comprehensive Westinghouse experience in the field of LWR pressure vessel and internals segmentation, and particularly in previous reactor internals segmentation project for Jose Cabrera NPP. This article shows the development of all the activities included: cutting method selection, preparatory works, cutting activities, waste characterization and packaging activities. (Author)

  10. Considerations on Improving Seismic Design at NPP Emergency Response Center after Fukushima Daiichi NPP Accident%福岛核事故后对我国核电厂应急控制中心抗震设防的思考

    付强; 陈晓秋; 岳会国; 潘蓉

    2011-01-01

    结合福岛核事故后对我国核电厂进行的核安全检查,分析了我国核安全法规关于核电厂应急控制中心的要求以及福岛核事故的经验教训,提出目前我国核电厂应急控制中心采用民用抗震设防标准进行抗震设防,无法保证在由地震引发的应急事故工况下应急控制中心的功能,应该适当提高其抗震设防级别。%Based on the nuclear safety impections at NPPs carried out in China in the aftermash of Fukushima Daiichi NPP accident, the requirements for NPP emergency response centers in the nuclear safety regulations and the lessons of the Fukushima Daiichi NPP accident are analylzed. The seismic design of the current NPP emergency response centers in China, based on the civil building seismic design code, is considered to be unable to ensure the function of emergency response centers in emergency accident conditions caused by earthquake, except as otherwise impporved.

  11. Adaptation of the electric system of Almaraz NPP ATEX (explosive atmospheres); Adaptacion del Sistema Electrico de la CN de Almaraz a la normativa ATEX (Atmosferas Exposivas)

    Felix, E.

    2014-07-01

    To comply with Royal Decree 681/2003, which assimilates Directive 1999/92/CE of the European Parliament and of the Council, ATEX Classified Zones were defined by Almaraz NPP, along with the equipment located in these areas that are possible sources of ignition. The facilities were then adapted accordingly, either by moving the equipment outside the classified zones or replacing it with ATEX marked equipment. The bases for choosing this equipment are briefly explained in this paper. (Author)

  12. Process for evaluation of renewal of the operating permit for Garona NPP.; Proceso para la evaluacion de la renovacion de la autorizacion de explotacion de Garona

    Zarzuela Jimenez, J.

    2009-07-01

    Process for evaluation of renewal of the operating permit for Garona NPP. The Santa Maria de Garona nuclear power plant has requested the renewal of its operating permit for a period of ten years, this implying extension of the operating lifetime of the facility beyond the 40 years originally established. This article explains the process of evaluation that the CN is carrying out in order to draw up a report on the technical feasibility of this proposal. (Author)

  13. Organization and management of maintenance in the NPP's Asco and Vandellos II; Organizacion y gestion del mantenimiento en las centrales nucleares Asco y Vandellos II

    Folguera, M.; Corral, A.

    2014-04-01

    The article starts with a description of the international framework that, using technical instructions, guides and guidelines, regulates the maintenance of nuclear power plants. It also outlines the characteristics of the organization and management of maintenance in the NPP's operated by ANAV. Such management is supported in a variety of processes and programs among which are: work management, training and qualification, operational experience, supervision, foreign material exclusion, work management in RP areas and outage preparation. (Author)

  14. Analysis of Alternatives for Dismantling of the Equipment in Building 117/1 at Ignalina NPP - 13278

    Poskas, Povilas; Simonis, Audrius [Lithuanian Energy Institute, Kaunas (Lithuania); Poskas, Gintautas [Lithuanian Energy Institute, Kaunas (Lithuania); Kaunas University of Technology, Kaunas (Lithuania)

    2013-07-01

    Ignalina NPP was operating two RBMK-1500 reactors which are under decommissioning now. In this paper dismantling alternatives of the equipment in Building 117/1 are analyzed. After situation analysis and collection of the primary information related to components' physical and radiological characteristics, location and other data, two different alternatives for dismantling of the equipment are formulated - the first (A1), when major components (vessels and pipes of Emergency Core Cooling System - ECCS) are segmented/halved in situ using flame cutting (oxy-acetylene) and the second one (A2), when these components are segmented/halved at the workshop using CAMC (Contact Arc Metal Cutting) technique. To select the preferable alternative MCDA method - AHP (Analytic Hierarchy Process) is applied. Hierarchical list of decision criteria, necessary for assessment of alternatives performance, are formulated. Quantitative decision criteria values for these alternatives are calculated using software DECRAD, which was developed by Lithuanian Energy Institute Nuclear engineering laboratory. While qualitative decision criteria are evaluated using expert judgment. Analysis results show that alternative A1 is better than alternative A2. (authors)

  15. Comparison on Heat of Hydration between Current Concrete for NPP and High Fluidity Concrete including Pozzolan Powders

    Noh, Jea Myoung; Cho, Myung Sug [KEPCO Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Nuclear power plant (NPP) concrete structures are exposed to many construction factors that lower the quality of concrete due to densely packed reinforcements and heat of hydration since they are mostly constructed with mass concrete. The concrete currently being used in Korean NPPs is mixed with Type I cement and fly ash. However, there is a demand to improve the performance of concrete with reduced heat of hydration and superior constructability. Many advantages such as improving workability and durability of concrete and decreasing heat of hydration are introduced by replacing cement with pozzolan binders. Therefore, the manufacturing possibility of high fluidity concrete should be investigated through applying multi-component powders blended with pozzolan binders to the concrete structure of NPPs, while the researches on properties, characteristic of hydration, durability and long-term behavior of high fluidity concrete using multi-component cement should be carried out. High fluidity concrete which is made using portland cement and pozzlonan powders such as fly ash and blast furnace slag has better properties on heat of hydration than the concrete currently in use for NPPs

  16. A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

    Martina Adorni

    2011-01-01

    Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

  17. Seismic dynamic analysis of Heat Exchangers inside of the Auxiliary Buildings in AP1000{sup T}M NPP

    Di Fonzo, M.; Aragon, J.; Moraleda, F.; Palazuelos, M.; San vicente, J. L.

    2011-07-01

    Seismic dynamic analysis was carried out for the Heat Exchangers (RNS-HR) located inside of the Auxiliary Building in AP 1000{sup T}M NPP. The main function of the RNS-HX is to provide shutdown reactor cooling. These equipment's are safety-related. So the seismic analysis was done using the methodology for Seismic Category I (SCI) structures. The most important topic is that the RNS-HX shall withstand the effects of the Safe Shutdown Earthquake (SSE) and maintain the specified design functions. for the analysis, two finite element models (FEM) were built in order to investigate the structural response of the couple system of building and equipment. The response spectra method was used. The floor response spectra (FRS) at the slab-wall connection were used as input Lateral seismic restrain was necessary to added in order to achieve the natural frequency of 33 Hz. The global structural response was obtained by means of the modal combination method indicated in the Regulatory Guide 1.92.

  18. Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRAC{sub R}T

    Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto, E-mail: prey@tecnatom.e, E-mail: jaruiz@tecnatom.e, E-mail: nrivero@tecnatom.e [Tecnatom S.A., Madrid (Spain)

    2011-07-01

    In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC{sub R}T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC{sub R}T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC{sub R}T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)

  19. The Use of System Codes in Scaling Studies: Relevant Techniques for Qualifying NPP Nodalizations for Particular Scenarios

    V. Martinez-Quiroga

    2014-01-01

    Full Text Available System codes along with necessary nodalizations are valuable tools for thermal hydraulic safety analysis. Qualifying both codes and nodalizations is an essential step prior to their use in any significant study involving code calculations. Since most existing experimental data come from tests performed on the small scale, any qualification process must therefore address scale considerations. This paper describes the methodology developed at the Technical University of Catalonia in order to contribute to the qualification of Nuclear Power Plant nodalizations by means of scale disquisitions. The techniques that are presented include the so-called Kv-scaled calculation approach as well as the use of “hybrid nodalizations” and “scaled-up nodalizations.” These methods have revealed themselves to be very helpful in producing the required qualification and in promoting further improvements in nodalization. The paper explains both the concepts and the general guidelines of the method, while an accompanying paper will complete the presentation of the methodology as well as showing the results of the analysis of scaling discrepancies that appeared during the posttest simulations of PKL-LSTF counterpart tests performed on the PKL-III and ROSA-2 OECD/NEA Projects. Both articles together produce the complete description of the methodology that has been developed in the framework of the use of NPP nodalizations in the support to plant operation and control.

  20. Evaluation of chemical phenomena that could have an effect on the performance of recirculation strainers in a Ringhals PWR; Bedoemning av kemiska fenomen med betydelse foer silfunktionen i Ringhals PWR-reaktorer

    Liljenzin, Jan-Olov [Liljenzins data och kemikonsult, Goeteborg (Sweden)

    2005-01-15

    An evaluation has been made of the various chemical phenomena that could have an effect on the performance of recirculation strainers after a LOCA in a PWR. Values of pH and concentrations in the water at the bottom of the containment have been calculated as functions of time and temperature for a postulated LOCA. The behaviour of glass wool insulation, its dissolution, and precipitation of amorphous silic acid have been evaluated. Also the corrosion of galvanized surfaces has been considered. Dissolution of zinc by hot boric acid solution can lead to a later precipitation of amorphous zinc hydroxide or phosphate when pH increases and temperature drops. Also a possible growth of microorganisms is discussed. A rough classification of the various phenomena possible along a simplified time scale yields the following conclusions: Hours after the beginning of the LOCA: Precipitation of zinc hydroxide and/or phosphate. Dissolution of glass wool giving rise to an increasing concentration of silic acid in the water. Days after the beginning of the LOCA: Continued dissolution of glass wool and increasing concentration of silica in the water. Perhaps a precipitation of phosphates or carbonates of the metal ions released during dissolution of glass wool. Weeks after the beginning of the LOCA: Continued slow dissolution of glass wool leading to a risk of precipitation of amorphous silica. Perhaps a precipitation of phosphates or carbonates of the metal ions released during dissolution of glass wool. Initial growth of microorganisms in the water and on surfaces after mutations and adaptation to the existing environment. Months after the beginning of the LOCA: Continued slow dissolution of glass wool leading to a risk of precipitation of amorphous silica. Perhaps a precipitation of phosphates or carbonates of the metal ions released during dissolution of glass wool. Continued growth of adapted microorganisms.