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Sample records for 300mwe pwr npp

  1. Core loading pattern optimization of a typical two-loop 300 MWe PWR using Simulated Annealing (SA), novel crossover Genetic Algorithms (GA) and hybrid GA(SA) schemes

    International Nuclear Information System (INIS)

    Highlights: • SA and GA based optimization for loading pattern has been carried out. • The LEOPARD and MCRAC codes for a typical PWR have been used. • At high annealing rates, the SA shows premature convergence. • Then novel crossover and mutation operators are proposed in this work. • Genetic Algorithms exhibit stagnation for small population sizes. - Abstract: A comparative study of the Simulated Annealing and Genetic Algorithms based optimization of loading pattern with power profile flattening as the goal, has been carried out using the LEOPARD and MCRAC neutronic codes, for a typical 300 MWe PWR. At high annealing rates, Simulated Annealing exhibited tendency towards premature convergence while at low annealing rates, it failed to converge to global minimum. The new ‘batch composition preserving’ Genetic Algorithms with novel crossover and mutation operators are proposed in this work which, consistent with the earlier findings (Yamamoto, 1997), for small population size, require comparable computational effort to Simulated Annealing with medium annealing rates. However, Genetic Algorithms exhibit stagnation for small population size. A hybrid Genetic Algorithms (Simulated Annealing) scheme is proposed that utilizes inner Simulated Annealing layer for further evolution of population at stagnation point. The hybrid scheme has been found to escape stagnation in bcp Genetic Algorithms and converge to the global minima with about 51% more computational effort for small population sizes

  2. Study and Economics Analysis for 18-month Refueling Management on Power Uprate of a 300 MWe NPP%30万千瓦核电厂提升功率下的18个月换料燃料管理研究及经济性分析

    Institute of Scientific and Technical Information of China (English)

    曹泓; 王丽华

    2013-01-01

    鉴于近年来提升功率在大量现役核电厂中的成功应用,以及长周期、高燃耗、低泄漏的堆芯燃料管理技术所带来的更高的燃料利用率,本文对现役30万千瓦核电厂基于提升功率下开展了长周期、高燃耗、低泄漏的堆芯燃料管理方案研究.本文初步设计的燃料管理策略在满足有关的设计准则和要求基础上,可满足堆芯额定热功率为1250 MW的18个月长周期的堆芯燃料管理目标.与现行的堆芯燃料管理方案相比,可使核电厂在提升功率的同时进一步提高燃料利用率,经济性得到显著提高.本研究作为一项技术支持和技术储备,对30万千瓦级核电厂的功率提升和燃料管理方案优化具有较强的指导意义.%In recent years,power uprate is successfully applied in many nuclear power plants.Moreover,a longer cycle,higher uprate burnup and lower leakage fuel management strategy could enhance the fuel utilization.Therefore,the purpose of this article is to study a longer cycle,uprate burnup and lower leakage fuel management for a 300MWe NPP after power uprate.The results show that the concluded fuel management scheme for a 300MWe NPP after power uprate achieves the projected 18-month refueling cycle design objectives with the nominal thermal power of 1 250 MW and meets the design criteria.As compared to the current fuel management strategy of a 300MWe NPP,the advanced strategy in present study gains a power uprate,enhances the fuel utilization and improves the operation economy.As a technical support and reserve,the study will provide significant instructions on power uprate of a 300MWe NPP and optimization of fuel management strategy.

  3. Steam Generator Chemical Cleaning Application: Korean Experience in PWR NPP

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power (KHNP) performed an EPRI/SGOG chemical cleaning of the secondary side of the steam generators at Ulchin Unit 3 (UCN3) in March 2011 and at Ulchin Unit 4 (UCN4) in September 2011. The steam generator chemical cleaning (SGCC) was performed with venting at the top-of-tube sheet (TTS) and at tube support plates (TSPs) 4, 5, 6, 7, 8, 9, and 10. A primary objective of this SGCC was to address outer diameter stress corrosion cracking (ODSCC), which has been observed at the TTS and TSPs in the UCN3 SGs. The EPRI/SGOG process has been shown to effectively reduce prevailing ODSCC rates at the TTS and TSPs, particularly when applied with periodic venting in this application. This was the first full-length SGCC campaign with venting performed in Korea. Ulchin Unit 3 commenced commercial operation in August 1998 and Ulchin Unit 4 commenced commercial operation in December 1999. UCN3 and UCN4 are a two-loop pressurized water reactor (PWR) of the Korea Standard Nuclear Plant (KSNP) design. The SGs contain high-temperature mill annealed (HTMA) Alloy 600 tubing and are similar in design to the Combustion Engineering CE-80. The KSNP SGs have been susceptible to outer diameter stress corrosion cracking (ODSCC), which is consistent with operating experience for other SGs containing Alloy 600HTMA tubing material. The UCN3/4 SGs have recently begun to experience ODSCC. Hankook Jungsoo Industries Co., Ltd (HaJI) was selected as the cleaning vendor by KHNP. To date, HaJI has completed five Advanced Scale Conditioning Agent (ASCA) cleaning applications and two EPRI/SGOG Steam Generator Chemical Cleaning (SGCC) campaigns for KHNP. The goal of total deposit removal of the applications were successfully achieved and the amounts are 3,579 kg at UCN3 and 3,786 kg at UCN4 which values were estimated before each cleaning by analysing ECT signal and liquid samples from the SGs. The deposits from the SGs were primarily composed of magnetite. There were no chemical

  4. NOx emission and thermal efficiency of a 300 MWe utility boiler retrofitted by air staging

    Energy Technology Data Exchange (ETDEWEB)

    Li, S.; Xu, T.M.; Hui, S.; Wei, X.L. [Chinese Academy of Sciences, Beijing (China). Inst. of Mechanics

    2009-09-15

    Full-scale experiments were performed on a 300 MWe utility boiler retrofitted with air staging. In order to improve boiler thermal efficiency and to reduce NOx emission, the influencing factors including the overall excessive air ratio, the secondary air distribution pattern, the damper openings of CCOFA and SOFA, and pulverized coal fineness were investigated. Through comprehensive combustion adjustment, NOx emission decreased 182 ppm (NOx reduction efficiency was 44%), and boiler heat efficiency merely decreased 0.21%. After combustion improvement, high efficiency and low NOx emission was achieved in the utility coal-fired boiler retrofitted with air staging, and the unburned carbon in ash can maintain at a desired level where the utilization of fly-ash as byproducts was not influenced.

  5. Probabilistic safety assessment of French 900 and 1,300 MWe nuclear plants

    International Nuclear Information System (INIS)

    Although reactor design is mainly governed by deterministic principles in France, the probabilistic approach has been considered an important aid to safety analysis since the early seventies. Various partial probabilistic studies have been performed by Electricite de France, by IPSN and by Framatome, for various types of reactor. In particular, these studies have made it possible to assess the reliability and availability of nuclear power plants safety systems as well as the probability of accident scenarios and have helped to define technical specifications (especially, allowed operating times in the event of a partial unavailability of safety systems). Simultaneously, evaluation methods and corresponding software have been widely developed. Besides. EDF has implemented the Systeme de Recueil de Donnees de Fiabilite - SRDF (Reliability Data Collection System) which allows follow-up of equipment behaviour on all the operating units, and has led to a particularly representative data base. In 1982 the decision was taken at IPSN to carry out a complete PSA for a standard reactor of the 900 MWe type, and in 1986 EDF launched an equivalent study on a 1,300 MWe reactor, taking Unit 3 Paluel as reference. These PSAs were terminated in the course of the first quarter of 1990

  6. NO{sub x} emission and thermal efficiency of a 300 MWe utility boiler retrofitted by air staging

    Energy Technology Data Exchange (ETDEWEB)

    Li, Sen; Wei, Xiaolin [Institute of Mechanics, Chinese Academy of Sciences, No.15 Beisihuanxi Road, Beijing 100080 (China); Xu, Tongmo; Hui, Shien [State Key Laboratory of Multiphase Flow in Power Engineering, Xi' an Jiaotong University, 28 Xian Ning Road, Xi' an 710049 (China)

    2009-09-15

    Full-scale experiments were performed on a 300 MWe utility boiler retrofitted with air staging. In order to improve boiler thermal efficiency and to reduce NO{sub x} emission, the influencing factors including the overall excessive air ratio, the secondary air distribution pattern, the damper openings of CCOFA and SOFA, and pulverized coal fineness were investigated. Through comprehensive combustion adjustment, NO{sub x} emission decreased 182 ppm (NO{sub x} reduction efficiency was 44%), and boiler heat efficiency merely decreased 0.21%. After combustion improvement, high efficiency and low NO{sub x} emission was achieved in the utility coal-fired boiler retrofitted with air staging, and the unburned carbon in ash can maintain at a desired level where the utilization of fly-ash as byproducts was not influenced. (author)

  7. Iodine speciation and behavior under normal PWR operating primary coolant conditions: analysis of thermodynamic evaluations and NPP feedback

    Energy Technology Data Exchange (ETDEWEB)

    Tigeras, A.; Bachet, M.; Catalette, H., E-mail: arancha.tigeras@edf.fr, E-mail: martin.bachet@edf.fr, E-mail: hubert.catalette@edf.fr [Electricite de France (France); Simoni, E., E-mail: simoni@ipno.in2p3.fr [Univ. Paris XI (France)

    2010-07-01

    Iodine is one of the most important fission products, in terms of nuclear reactor safety, due to its high fission yield, significant radiobiological hazard, and potential volatility. The iodine environmental and biological risks have been extensively studied in case of severe reactor accidents. Nevertheless, very little information is available about iodine behavior under normal PWR operating conditions. The work reported in this paper intends to enrich existing knowledge about iodine species behavior (I{sup -,}, I{sub 3} , I{sub 2}, HOI, IO{sup -}) at full power, transient periods (power reductions, depressurizations) and during shutdowns. For this purpose, thermodynamic calculations were conducted, and their results were compared with previous predictions offered by other authors and with the experimental data provided by NPP. In the highlight of the thermodynamic calculations and NPP feedback it is concluded that the iodine speciation depends basically on redox potential and water radiolysis phenomenon: The experimental values confirm that the iodine ionic form I- is the preponderant specie during normal operation (I{sub 2}<2%) and shutdowns (I{sub 2}<9%); During shutdowns: High punctual [I{sub 2}] (20-40%) can be observed in presence of fuel failures, following the iodine spike during power or pressure variations. The fuel oxidation by radiolysis products can lead to I{sub 2} formation inside the gap and its subsequent release through the cladding defect; Once in the primary coolant, the I{sub 2} is transformed into I{sup -} or IO{sub 3}{sup -}/IO{sub 4}{sup -}, depending on the water oxidation conditions; and, The [Li] and the primary coolant temperature seem to be variable with a secondary influence on iodine speciation while the existence of redox potential threshold appears as main key to control the control formation of volatile/ non-volatile iodine forms. This paper summarizes the major results of iodine thermodynamic studies and PWR feedback permitting

  8. Iodine speciation and behavior under normal PWR operating primary coolant conditions: analysis of thermodynamic evaluations and NPP feedback

    International Nuclear Information System (INIS)

    Iodine is one of the most important fission products, in terms of nuclear reactor safety, due to its high fission yield, significant radiobiological hazard, and potential volatility. The iodine environmental and biological risks have been extensively studied in case of severe reactor accidents. Nevertheless, very little information is available about iodine behavior under normal PWR operating conditions. The work reported in this paper intends to enrich existing knowledge about iodine species behavior (I-,, I3 , I2, HOI, IO-) at full power, transient periods (power reductions, depressurizations) and during shutdowns. For this purpose, thermodynamic calculations were conducted, and their results were compared with previous predictions offered by other authors and with the experimental data provided by NPP. In the highlight of the thermodynamic calculations and NPP feedback it is concluded that the iodine speciation depends basically on redox potential and water radiolysis phenomenon: The experimental values confirm that the iodine ionic form I- is the preponderant specie during normal operation (I2<2%) and shutdowns (I2<9%); During shutdowns: High punctual [I2] (20-40%) can be observed in presence of fuel failures, following the iodine spike during power or pressure variations. The fuel oxidation by radiolysis products can lead to I2 formation inside the gap and its subsequent release through the cladding defect; Once in the primary coolant, the I2 is transformed into I- or IO3-/IO4-, depending on the water oxidation conditions; and, The [Li] and the primary coolant temperature seem to be variable with a secondary influence on iodine speciation while the existence of redox potential threshold appears as main key to control the control formation of volatile/ non-volatile iodine forms. This paper summarizes the major results of iodine thermodynamic studies and PWR feedback permitting to suggest some possible recommendations, intended for inclusion in NPP guidelines

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

    Science.gov (United States)

    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.

  10. Investigation on Interim Storage of Spent Fuel for PWR NPP%压水堆乏燃料中间贮存技术研究

    Institute of Scientific and Technical Information of China (English)

    刘彦章; 王鑫; 袁呈煜; 莫怀森

    2015-01-01

    The spent fuel interim storage and treatment status of pressurized water reactor in main nuclear power countries was investigated and the recent trends in the spent fuel interim storage of the pressurized water reactor were analyzed. Dry storage technology of nuclear spent fuel will be the main stream of interim storage for future PWR spent fuel storage. Suggestions were made for nuclear spent fuel storage and processing combined with the status of China's nuclear spent fuel in pressurized water reactor.%本文通过调研主要核电国家的压水堆核电站乏燃料中间贮存与处理现状,分析研究近年来在压水堆核电站乏燃料中间贮存方面的趋势,明确乏燃料干式贮存技术将是未来压水堆核电站乏燃料中间贮存的主流。结合我国压水堆核电站乏燃料的现状并对未来核电站乏燃料贮存与处理工作提出建议。

  11. Study on Concentrating Treatment Test of Simulated Radioactive Wastewater Containing Boron by Reverse Osmosis Membrane in PWR NPP%压水堆核电厂模拟含硼废液反渗透浓缩试验研究

    Institute of Scientific and Technical Information of China (English)

    叶欣楠; 姜百华; 范雯雯; 张志银; 严沧生

    2015-01-01

    采用中试规模反渗透试验装置,在浓水全回流的运行模式下,研究了反渗透系统在压水堆核电厂放射性废液处理中的应用。重点考察了该系统对模拟废液中硼的截留效果,并进一步研究了反渗透水处理工艺对模拟放射性核素的截留效果。结果表明,海水型聚酰胺复合膜对原水中硼的截留率可达83.3%以上,并将原水中硼浓度浓缩至10000 mg/L以上。试验结果同时表明,上述试验装置对于核素如钴和铯的截留率可达97.9%以上。%The reverse osmosis membrane equipment in PWR NPP was employed to investigate the application of pilot scale system in the radioactive wastewater treatment at the full recirculation operation .The removal performance of the equipment for the boron and the radioactivity nuclide were studied ,respectively .The experimental results show that the removal efficiency of the aromatic polyamide composite reverse osmosis membrane for boron is over 83.3% and the concentration of boron in concentrate is over 10 000 mg/L .The experimental results also show that the removal efficiency of two nuclides including cobalt and cesium is over 97.9% .

  12. Study on severe accident mitigation measures for the development of PWR SAMG

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In the development of the Severe Accident Management Guidelines (SAMG), it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR), Station Blackout (SBO), and Anticipated Transients without Scram (ATWS) in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management (SAM) such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.

  13. 压水堆核电站堆芯物理/热工水力耦合特性研究%Investigation on Coupling Characteristics of Neutronics/Thermal-hydraulics of PWR NPP Core

    Institute of Scientific and Technical Information of China (English)

    郑勇; 彭敏俊; 夏庚磊; 刘新凯

    2014-01-01

    采用RELAP5‐HD作为堆芯耦合计算程序,以秦山核电二期工程反应堆堆芯为研究对象,建立堆芯活性区的物理/热工水力耦合模型,在此基础上进行了稳态计算和掉棒事故仿真研究。结果表明,使用RELAP5‐HD计算得到的结果与电厂实测值符合较好,获得的掉棒事故参数曲线能准确反映事故工况下的参数变化趋势。稳态和事故工况的计算结果均符合堆芯物理/热工水力反馈效应的理论分析,证实了所建立的堆芯耦合模型的准确性,为下一步进行核电站系统的仿真分析提供基础。%In this paper ,an integrated neutronics/thermal‐hydraulic model for the reac‐tor of Qinshan Phase Ⅱ NPP project was developed ,using the RELAP5‐HD as core coupled computational code .Based on the coupled model ,the steady state calculation and the rod drop transient simulation were performed .The results show that the values obtained from RELAP5‐HD calculation agree well with the available measured data ,and the calculated accident curves can predict all major parameters trends of the transient with good accuracy .Both steady state and transient calculation results are in accordance with the theoretical analysis from the feedback aspect of coupled reactor neutronics/thermal‐hydraulics ,this demonstrates that a successful coupled model of Qinshan PhaseⅡ NPP core has been developed ,and the established model provides a good foundation for further simulation analysis of the nuclear power plant system .

  14. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    International Nuclear Information System (INIS)

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  15. Shielding and Criticality Safety Analysis of KSC-1 Cask for the High Burnup PWR Spent Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Jang, Jung Nam; Hwang, Yong Hwa; Kwon, In Chan; Min, Duck Kee; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    KSC-1 (KAERI Shipping Cask-1) was designed and manufactured with a pure domestic technology in 1985 in order to transport a PWR spent fuel assembly from nuclear power plant to PIEF (Post-Irradiation Examination Facility) of KAERI. Since the first transportation of the fuel assembly from Kori-1 NPP was carried out by the cask in 1987, 19 shipments for the PWR spent fuels have been done successfully by now. Maximum discharge burnup of PWR in Korea has been extended from the late 1990s in order to reduce the cost of power generation. From this cause, allowable design values of the initial enrichment and the cooling time for the cask have been changed three times: year 2003, 2007 and 2010. Radiation shielding and criticality of KSC-1 were analyzed for all the PWR fuel type irradiated in Korea NPP to renew the design approval

  16. Maintenance centered on reliability applied to a NPP auxiliary feedwater system

    International Nuclear Information System (INIS)

    The main objective of maintenance in a NPP is to assure that structures, systems and components will perform their design functions with reliability and/or availability in order to allow a safe and economic electric power generation. Reliability-Centered Maintenance (RCM) is a method of systematic review to either develop or optimize Preventive Maintenance Programs. This paper presents the objectives, concepts, organization and methods used in the application of RCM to NPP. Some application examples are include in this paper, considering some components of the Auxiliary Feedwater System of a generic Westinghouse designed two-loop PWR NPP. (author). 4 refs., 3 figs

  17. Construction prospects of new power units at Khmelnitskij NPP site

    International Nuclear Information System (INIS)

    According to the Energy Strategy of Ukraine for a period up to 2030 it is planned to put into operation power units 3 and 4 of Khmelnitskij NPP by year 2016. In this work considerations are presented on the possible options while selecting reactor unit type for Khmelnitskij NPP power units 3 and 4, which is the main determinant of the cost, construction and commissioning time, and utilization of the existent civil structures. To optimize Khmelnitskij-3 and 4 construction, a survey of the data has been conducted with regard to the possibility of construction of new power units of PWR/VVER type at Khmelnitskij NPP site. The multivariable analysis has been performed based on the projects technical and cost data, construction time and conditions, as well as their compliance with the IAEA and EUR safety requirements for new power units. (author)

  18. 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

    Institute of Scientific and Technical Information of China (English)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Design of an FPGA-based PWR ATWS mitigation system

    International Nuclear Information System (INIS)

    The present research is to explore the feasibility and conceptual design by using triple-redundant FPGA-based system for Anticipated-Transient-Without-Scram (ATWS) Mitigation System and Actuation Circuit (AMSAC) of a pressurized water reactor (PWR) type nuclear power plant (NPP). The Taipower's (Taiwan Power Company) Maanshan NPP was chosen for demonstration. An engineering simulated interface between AMSAC system and reactor/plant systems of Maanshan NPP was developed to provide an environment to validate the triple-redundant FPGA-based system. The software-free FPGA-based nuclear instrumentation and control (I and C) systems can easily be used for the modernization of the Taipower's nuclear power plant analog systems, thus may reduce the safety risk of undetectable software faults and common cause failures, and also minimize the regulatory licensing efforts and cost. (author)

  1. Foreign NPP decommissioning

    International Nuclear Information System (INIS)

    Different versions of NPP decommisioning, which worked out their life are considered. Dismantling work technology as well as devices for cutting and decontamination of equipment and concrete structures are described. Data on the quantity of shutdown and dismantled NPPs are given. It is noted that to perform successfully dismantling works it is necessary to: choose NPP decommisioning version; calculate radioactivity level; substantiate necessity of decontamination; develop the plan of removal of radioactive equipment; radioactive concrete and structures; contaminated systems; transport and bury solid, liquid and gaseous radioactive and chemical wastes; evaluate the accepted solutions of dismantling from the point of view of the effect on environment; determine costs. It is shown that optimal period of complete or partial dismantling after the NPP decommisioning is 15 years. NPPs dismantling expenditures can reach 10-15% of expenditures for their construction

  2. IAEA activities on safety aspects of NPP ageing

    International Nuclear Information System (INIS)

    A review of IAEA activities concerned with safety aspects of nuclear power plants ageing is given for the period from 1995 to 1998 with the prospects till year 2000. Coordinated Research programs were conducted on Management Ageing of Concrete Containment Buildings; Management of Ageing of In-Containment I and C cables. TECDOCs were published on Assessment and Management of Ageing of Major NPP Components Important for Safety of CANDU, PWR and BWR NPPs. Technical Committee Meetings and Interregional training courses concerned with the same subjects were held

  3. Survey of Water Chemistry and Corrosion of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ki Sok; Hong, Bong Geon

    2008-06-15

    Status of water chemistry of nuclear power plant and materials corrosion has been surveyed. For PWR, system chemistry of primary coolant and secondary coolant as well as the related corrosion of materials was surveyed. For BWR, system chemistry as whole has been surveyed with its accompanying corrosion problems. Radiolysis of coolant water and activation of corrosion products also was surveyed. Future NPP such as supercritical water cooled reactor and fusion reactor has also been surveyed for their water chemistry and corrosion problems. As a result, proposal for some research items has been suggested. Some related corrosion research techniques and electrochemical fundamentals are also presented.

  4. Two managerial grids in NPP

    International Nuclear Information System (INIS)

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

  5. Brazilian system for NPP personnel training

    International Nuclear Information System (INIS)

    Brazil has two nuclear power plants in operation (Westinghouse design 657 MW PWR-type Angra-1 NPP, in commercial operation since 1985), (Siemens/KWU-design 1300 MW PWR-type Angra-2 NPP in commercial operation since 2000) and a third for which about 80% of components are already stored on site, awaiting for construction (Siemens/KWU-design 1300 MW PWR-type Angra-3 NPP). The training center is operated by ELETRONUCLEAR at Angra site. It is divided in main training center, simulator training center and maintenance training center. All training activities for Angra-1 and 2 operation personnel are performed in the training center, with exception of simulator training for licensed control room operators of Angra 1, which is performed at training centers in the U.S. and Spain. International cooperation and assistance have been extensively used during last years . All training modules are developed and updated by utility staff. Most of Instructors come from the Operations staff. Training methodology is characterized by modules which follow international practices. The simulator training center was built in 1985 and houses a full-scope training simulator used as design reference is the one used for Angra-2 NPP. Since 1985 an extensive scope of courses for operators, managers and other specialists from a total of eleven NPPs and other organizations in Germany, Spain, Argentina and Switzerland is being provided by ELETRONUCLEAR with utilization of the simulator and Angra-2 training center staff. Provision of such courses for NPP operators from other countries results in acquisition, by ELETRONUCLEAR instructors, of considerable experience in the field of training of NPP operation personnel. This experience is extremely useful for qualification of Angra-2 operators. Maintenance training center started its activities in July 1996. It presently consists of classrooms, meeting rooms and offices for staff. A maintenance workshop is currently under construction and is going to be

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  7. Reconsidering the site requirements for NPP on Olt River

    International Nuclear Information System (INIS)

    Site studies for CANDU type NPP began in a careful manner since 1982 as a first part of the Nuclear Power Plant Romanian Program adopted by political and governmental authorities at the time. A team was charged to develop all packages of the necessary main studies. The first Romanian NPP CANDU 6 type reactor gone to erection on Cernavoda site, planned to have 5 units and, like Wolsong NPP, applied the same design for the nuclear island. For the BOP parts the ANSALDO-GE project was applied with a thorough concern about requirements raised by connection to NSP. The first mission of design and research multi-branch team was to adapt the NPP Cernavoda project having an open water cooling circuit 'once-through' to the new parameters of a close recirculation water cooling circuit. Also, the structural design was re-evaluated for the case of soft foundation strata instead of hard rock ones. The close recirculation water cooling circuit system was applied for PWR NPP type like in French or other nuclear projects where a rich water source was not available. In case of CANDU type projects cooling water loops were not built so far. The close recirculation circuit of water cooling implies other parameters of the cooled systems and for turbine steam as well, needing very large cooling towers. The initial Romanian nuclear program implied the construction of a 4 units NPP sited near Olt river. This river runs in Transylvania region of Romania from east to west near Boitza village on the northern side of Fagaras Mountains. From geological and geophysical points of view the following main characteristics were found by surveying the Olt River valley: there exist two faults having about east-west direction, one of these having small seismic activities; the stratum for foundation consists of marls or sandy marls; there exist also underground small bags of natural gas or salty strata here and there, as detected by geotechnical borings near the Olt river. The average multiannual water

  8. Control of NPP aging processes

    International Nuclear Information System (INIS)

    The concept of the control program on the NPP aging processes is considered. The methodological algorithms for the working programs plotting and realization, intended for accomplishing the measures for mitigating the aging mechanisms and factors, effecting the NPP safety, are presented. The efficiency of the equipment for the aging processes control and power units systems, aimed at the control of the NPP service life, is analyzed

  9. Performance of PWR study in the technology supplier countries: south korea and japan case

    International Nuclear Information System (INIS)

    Electricity is needed as an infrastructure to support the national economic growth. For economic development sustainability, energy alternatives should be provided. Nuclear Power Plant (NPP) become the alternative of electricity generation for optimum energy mix in Indonesia and planned to operate in the 2016. Several studies have already done to prepare the NPP construction. This study focused on NPP performance especially PWR type in Asia, namely Japan and South Korea. Methodology used in this is literature tracing and a small calculation. The energy availability per unit per year is used as a parameter for evaluating the NPP performance. This conclusion are 1) the amount of NPP - PWR type in Japan is 22 units with total operational experiences 526 reactor-years and the average energy availability factor about 70.7% per unit per year. Meanwhile for the same type South Korea has 16 unit with total operational experience 222 reactor-years and average availability factor per unit per year is about 86.9%. 2) the 1000 class of PWR type both South Korea and Japan have 14 units. The operational experiences for thi class is 170 reactor-year for South Korean and 307 reactor-year for Japan. Meanwhile the average availability factor per unit per year is about 87.0% for South Korea and 69.6% for Japan. 3) the average availability factor is closed to capacity factor, so is important for real figure in assuming the techno-economic parameters, because it will influence the result o economic calculation. (author)

  10. Preparation for Ignalina NPP decommissioning

    International Nuclear Information System (INIS)

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

  11. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    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

  12. PWR type reactor

    International Nuclear Information System (INIS)

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  13. NPOESS Preparatory Project (NPP) Science Overview

    Science.gov (United States)

    Butler, James J.

    2011-01-01

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

  14. Psychology of NPP operation safety

    International Nuclear Information System (INIS)

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

  15. Calculation of economic and financing of NPP and conventional power plant using spreadsheet innovation

    International Nuclear Information System (INIS)

    The study for calculating the economic and financing of Nuclear Power Plant (NPP) and conventional power plant using spreadsheet Innovation has been done. As case study, the NPP of PWR type of class 1050 MWe is represented by OPR-1000 (Optimized Power Reactor, 1000 MWe) and the conventional plant of class 600 MWe, is coal power plant (Coal PP). The purpose of the study is to assess the economic and financial feasibility level of OPR-1000 and Coal PP. The study result concludes that economically, OPR-1000 is more feasible compared to Coal PP because its generation cost is cheaper. Whereas financially, OPR-1000 is more beneficial compared to Coal PP because the higher benefit at the end of economic lifetime (NPV) and the higher ratio of benefit and cost (B/C Ratio). For NPP and Coal PP, the higher Discount Rate (%) is not beneficial. NPP is more sensitive to the change of discount rate compared to coal PP, whereas Coal PP is more sensitive to the change of power purchasing price than NPP. (author)

  16. PWR degraded core analysis

    International Nuclear Information System (INIS)

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  17. Trillo NPP full scope replica simulator project: The last great NPP simulation challenge in Spain

    International Nuclear Information System (INIS)

    In the year 2000, Trillo NPP (Spanish PWR-KWU design nuclear power plant) and Tecnatom came to the agreement of developing a Trillo plant specific simulator, having as scope all the plant systems operated either from the main control room or from the emergency panels. The simulator operation should be carried out both through a control room replica and graphical user interface, this latter based on plant schematics and softpanels concept. Trillo simulator is to be primarily utilized as a pedagogical tool for the Trillo operational staff training. Because the engineering grade of the mathematical models, it will also have additional uses, such as: - Operation engineering (POE's validation, New Computerized Operator Support Systems Validation, etc).; - Emergency drills; -Plant design modifications assessment. This project has become the largest simulation task Tecnatom has ever undertaken, being structured in three different subprojects, namely: - Simulator manufacture, Simulator acceptance and Training material production. Most relevant technological innovations the project brings are: Highest accuracy in the Nuclear Island models, Advanced Configuration Management System, Open Software architecture, Human machine interface new design, Latest design I/O system and an Instructor Station with extended functionality. The Trillo simulator 'Ready for Training' event is due on September 2003, having started the Factory Acceptance Tests in Autumn 2002. (author)

  18. Temelin NPP status and challenge

    International Nuclear Information System (INIS)

    In this presentation author deals with the NPP Temelin status and challenge. It was concluded that: - Temelin NPP was modified from the every beginning in order to meat internationally acceptable safety level; - IAEA, US and Western countries safety principles, criteria and requirements are mostly applied; - Number of international safety review missions confirmed this fact; All assessments of the Temelin NPP have been positive and all recommendations were carefully considered and either implemented or other equivalent solution was found. Temelin NPP Halliburton NUS audit in 1992 stated that Temelin can be licensable, but licensibility could not be assured unless the audit team's technical and programmatic recommendations are implemented. ENCONET Consulting (Austria) in 1998 stated that: - After modifications are fully implemented, Temelin NPP will be a much safer plant than originally designed and much more safer than some of the already operating WWER 1000 plants; - The process of compatibility was specifically assured by selecting prudent practices acceptable in the Western countries. IAEA mission on Safety issues resolution (1996) stated that: - It is recognized that the Czech Electric Company (CEZ) has made a large effort to improve the design of Temelin independently of the identification of safety issues by the IAEA

  19. Regulatory aspects of NPP safety

    International Nuclear Information System (INIS)

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

  20. Ultrasonic measurement on RPV stud-bolt loading under hot transient of Qinshan NPP

    International Nuclear Information System (INIS)

    It is a continuation of research work for sealing analysis and tests on the PRV of PWR. It expounds that the key of solving thermal transient sealing problem lies in giving the thermal increment of stud-bolt fatigue life and transient loading spectrum for vessel analysis. The authors recounted the fundamental works and main results of ultrasonic measurement on RPV stud-bolt loading on the reactor of Qinshan Nuclear Power Plant. The measuring capability exceeds 1 m length and 300 degree C temperature. Therefore, it is possible to be used in the field of NPP

  1. Manufacture of nuclear fuel elements for commercial PWR in China

    International Nuclear Information System (INIS)

    Yibin Nuclear Fuel Element Plant (YFP) under the leadership of China National Nuclear Corporation is sole manufacturer in China to specialize in the production of fuel assemblies and associated core components for commercial PWR nuclear power plant. At the early of 1980's, it began to manufacture fuel assemblies and associated core components for the first core of QINSHAN 300 MW nuclear power plant designed and built by China itself. With the development of nuclear power industry in China and the demand for localization of nuclear fuel elements in the early 1990's, YFP cooperated with FRAMATOME France in technology transfer for design and manufacturing of AFA 2G fuel assembly and successfully supplied the qualified fuel assemblies for the reloads of two units of GUANGDONG Da Ya Bay 900 MW nuclear power plant (Da Ya Bay NPP), and has achieved the localization of fuel assemblies and nuclear power plants. Meanwhile, it supplied fuel assemblies and associated core components for the first core and further reloads of Pakistan CHASHMA 300 MW nuclear power plant which was designed and built by China, and now it is manufacturing AFA 2G fuel assemblies and associated core components for the first core of two units of NPQJVC 600 MW nuclear power plant. From 2001 on, YFP will be able to supply Da Ya Bay NPP with the third generation of fuel assembly-AFA 3G which is to realize a strategy to develop the fuel assembly being of long cycle reload and high burn-up

  2. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

  3. Current status of research and development of nuclear fuel elements for PWR in Indonesia

    International Nuclear Information System (INIS)

    Full text: The energy need of Indonesia is increasing due to the population growth and for the economic progress. The government of Indonesia intends to apply an optimum energy mix comprising all viable prospective energy sources. The Government Regulation No. 5 year 2006 indicates the target of energy mix until 2025 and the share of nuclear energy is about 2% of primary energy or 4% of electricity (4000 MWe). The first two units of NPP is expected to be operated before 2020 as stated in Act No. 17 year 2007 on National Long Term Development Planning 2005-2025. The first NPP to be operated in Indonesia is PWR type with capacity of 1000 MWe/unit. One of the strategies to strength and increase national capacity in the program for NPP introduction is domestication industry for nuclear fuel. To reach this purpose, the activities of research and development is focus on nuclear fuel production technology for PWR. Currently research and development activity in Indonesia is to produce prototype of nuclear fuel element for PWR in the form of test fuel pin or mini pin. In this paper we will presenting the pelletization and fabrication technology development. The existing facility was designed for PHWR fuel element of CIRENE type. Development of pelletization technology is carried out by modifying the compacting machine. Parameters of compacting and sintering are determined based on both compressibility and compactibility of the pellet as indicated by density and mechanical strength of the UO2 green pellets. The sintering parameters to be determined are temperature, heating rate, and soaking time. Currently, the fabrication process is under experiment. All of the data resulted from the experiment that will be presented in this meeting. (author)

  4. Full System Decontamination (FSD) with the CORD{sup R} Family prior to Decommissioning - Experiences at the German NPP Obrigheim 2007

    Energy Technology Data Exchange (ETDEWEB)

    Topf, Christian [Department STC-G, AREVA NP GmbH, P.O. Box 1109, 91001 Erlangen (Germany)

    2008-07-01

    Minimizing personal radiation exposure and obtaining material free for release are the highest priorities for the decommissioning of a NPP. This calls for a FSD as the first and most effective measure. AREVA NP has a long experience with FSDs, not only for operating plants but also for decommissioning in particular. Starting 1986 with the first decontamination for decommissioning at the NPP FR2, a German research reactor, AREVA NP has performed more than 10 FSDs worldwide prior to NPP dismantling. Based on these long term decontamination experiences including the successful performance of the FSD at the German PWR in Stade 2005, AREVA NP received the contract for the FSD prior to decommissioning at the German PWR in Obrigheim (357 MWe). The NPP Obrigheim was permanently shut down in May 2005 after 36 years of operation. The decontamination of the complete primary circuit and the auxiliary systems RHR and CVCS was performed in the first quarter of 2007. The total system volume was 160 m{sup 3}, total system surface approximately 8100 m{sup 2}. Decontamination was carried out with the worldwide approved decontamination process HP/CORD D UV, using NPP own systems and the AREVA NP AMDA decontamination equipment. In the paper detailed results of the decontamination will be outlined. An important value for further decommissioning activities was the remarkable dose rate reduction at the heavy components, especially the steam generators. The average decontamination factor achieved at the systems exceeded the value of 600. (author)

  5. The typical aging problem of the main component materials in PWR nuclear steam supply system

    International Nuclear Information System (INIS)

    The aging problem of equipment, components, and materials, is a systematic problem which exists throughout the whole process of NPP from design to retirement .In this paper, the basic information and current status of aging research and aging management is introduced, with detailed information about the typical aging problem and research work of main component materials in PWR nuclear steam supply system. Systematical studies have shown that: 1) the aging management work of NPP in China has started a little late and the research foundation is relatively weak; 2)the technology system and management system of aging management and lifetime assessment has not yet been formed; 3) the research work about aging problem of component materials concentrates only on the design verification except for the reactor pressure vessel , and there's no specialized research work on aging problem; 4)there's a lack of important data on aging evaluation and lifetime enhancement analysis . A systematical research work on component aging problem is suggested. (authors)

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

    DEFF Research Database (Denmark)

    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 phys...... 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....

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

    DEFF Research Database (Denmark)

    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....

  8. Risk indicators at Cofrentes NPP

    International Nuclear Information System (INIS)

    Following the tend to try to find indicators to show the excellence in the performance where Nuclear Power Plants are currently involved, Cofrentes NPP are managing several indicators related with risk. The concept of risk is classically associated with the product RISK = PROBABILITY * DAMAGE So what a risk based indicator will show is the probability of having a 'damage'. Speaking about a period of time, we will have frequencies of having 'damages'. What is call 'damage' can be differently interpreted depending of what we concern. In western NPP is very extended the concept of 'core damage', meaning the loss of fuel integrity, as a final state to avoid. This have carried in most of western NPP to develop a Probabilistic Risk Assessment (PRA/PSA), that using technical based in fault trees and event trees models, looks for the frequency to reach core damage. The PSA in Cofrentes NPP has been deeply applied to find weakness in the design and procedures, prioritizations in maintenance activities, quality assurance requirements, justifications to continued operation, and others. A Risk Monitor based in PSA models (and so monitoring the Core Damage Frequency) has been developed and is currently installed in the Control Room to help operators to control the risk associated with each configuration of availability or unavailability of equipments. This PSA Monitor is the source for some indicators that Cofrentes NPP has defined and are sharing with IAEA trying to find an standard. Maximum Core Damage Frequency reached and accumulated annual probability is calculated and compared with expected values and with predefined limits. As the PSA in Cofrentes NPP is only for at power Operations, there has been developed a methodology based on NUMARC 91-06 to measure and control the risk during shutdowns. The 'damage' here is a concept related with the safety functions. Some coefficients are applied to each configuration according with how the safety functions are fulfilled (defense

  9. Artificial intelligence and NPP safety

    International Nuclear Information System (INIS)

    The main tasks of the software for probabilistic safety analysis, thermal hydraulic analysis and probabilistic risk assessment are discussed. Their combination for direct improvement of NPP operation through information support of the staff is stressed. The general philosophy (in-depth protection) of computerized Emergency Response Guidelines (ERGs) - symptom based (safety parameters) and events-oriented (types of accident) is pointed out. The use of expert systems for proper diagnosing of the accident, its forecasting and finding the way of overcoming it is shown. Mandatory components of the modern management policy in abnormal situations are: the ERGs, the installation of Safety Parameter Display panels, the availability of an safety engineer (superviser); local, regional and national systems for monitoring of the radiation environment within and outside the NPP; local protection centres for maintenance in the case of accident. The importance of verification and validation (V and V) approach and benchmark exercise is stressed. Some peculiarities of the on-going implementation of the computerized information system for radiation control in Kozloduj NPP are discussed. 3 figs, 7 refs

  10. Ignalina NPP: living and working conditions

    International Nuclear Information System (INIS)

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

  11. NPP technical and economical parameters and safety

    International Nuclear Information System (INIS)

    The problem of the ratio between technical and economic indices of NPP and its safety has been considered. It is suggested that safety indices of NPP projected should be made allowance for, when calculating net cost of electric power generated so, that NPP with higher safety indices remained competitive. The problem can be solved using a special invariance fund for compensating the costs of protection measures taken. The amount of contribution is to be the higher, the lower are safety indices of NPP. 2 refs

  12. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  13. Physics of hydride fueled PWR

    Science.gov (United States)

    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.

  14. Emergency preparedness at Barsebaeck NPP in Sweden

    International Nuclear Information System (INIS)

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

  15. Distinct characteristics of NPP HRD and establishment of KINGS in Korea

    International Nuclear Information System (INIS)

    Full text:Korean government set-up nuclear energy department within the ministry of education in 1956 and joined IAEA in 1957 and set up nuclear energy agency in 1959, and installed the first research reactor in 1962. The Korean Government started constructing NPP in 1971 that had started commercial operation in 1978. The first oil shock in 1973 had devastated Korean economy and that made Korean Government to accelerate the construction of NPPs. Since then Korea steadily constructed NPP as well as invested in the development of indigenized NPP technology. During 1990s, Korea developed KSNP PWR 600 MWe NPP and in the last decade Korea developed APR1400 MWe NPP. Through the time, the engineers and operators involved in every field of nuclear industry is getting old and started to retire. Someone freshly out from the university with bachelor or post graduate degree will take many years to be able to understand how things running and operating in nuclear industry. Even in many years of job assignment, one cannot experience and understand all aspect of nuclear industry. It is this reason to establish a special educational system to teach people already in the field and to be able to see the whole picture by systematically teaching most of the related subject. In order to prevent any influence from existing university system, it was determined to establish KINGS (KEPCO International Nuclear Graduate School) as separate and independent institution and as a post-graduate institution. The curriculum of KINGS was set up along this philosophy, and has only one academic department, for example NPP Engineering Department, to make more interactions among faculty and students. Also the curriculum is set up to teach practical experience; hence the graduates can bridge between industry and academia as well as fill in the large gap of technical experience of older generation. Also another aim is to make KINGS international institution to share experience of Korean NPP development

  16. Reviewing NPP Cernavoda site evaluation

    International Nuclear Information System (INIS)

    The Nuclear Power Plant Cernavoda site was selected before the IAEA Safety Guide issue, during NUSS program development. The Romanian codes issued in 1976, as a regulatory body requirements, establish general criteria regarding safety concept and concentration limits of different radionuclides in air and water body and limits of individual or collective dose. In 1979 the Romanian Authority signed the contract with AECL to improve the CANDU-600 concept in the nuclear development programme and erection of 4 units on the Cernavoda site. The construction work started in 1980. In 1983 the former Romanian Government decided to build up another unit (finally it will be 5 units) on Cernavoda site, so the total gross electrical power we have 3,500 MW. The Canadian safety and quality standards or requirements was harmonized with the Romanian rules and regulations. Many studies, investigations and research were done to qualify the site and have a good knowledge about its characteristics coupled with CANDU-600 performance. The new evolution of the site was performed by Romanian technical staff in CITON and the final conclusions were favourable for erection and operation of NPP. The first unit of Cernavoda NPP is on operation and now the efforts are concentrated to continue the works for the unit 2. The paper underlines how the Cernavoda site characteristics meet IAEA Code of Practice and Safety Guides issued until now. (author)

  17. The decommissioning NPP A-1

    International Nuclear Information System (INIS)

    Project of decommissioning NPP A-1 is split into 4 main groups of tasks. Tasks in group 1 are focused on the solution of selected problems that have immediate impact on the environment. It is mainly the solution of problems in the building of cleaning station of wastage water and in the building with underground storage tanks for wastage water and solid radwaste, including the prevention of wash-out and penetration of contaminated soil from these buildings into surface and underground waters. A part of addressing these tasks is a controlled of generated radwaste-predominatly sludge with various physical and chemical properties. Tasks in group 2- following the removal of spent fuel-are focused on the management of all radwaste in the long-term storage facility, in the short-term storage facility, equipment of transport and technology part, equipment in hot cells. Tasks in group 3 are focused on development of technology procedures for treatment and conditioning of sludge, contaminated soils and concrete crush, saturated ionexes and ash from incineration facility of the Bohunice radwaste treatment and conditioning complex. Tasks in group 4 are focused on the methodology. And technical support for particular activities applicable during decommissioning NPP

  18. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    The authors believe the next generation nuclear power plant should be characterized by: (1) simplicity of design; (2) ease of operation and maintenance; (3) economic conformance with safety requirements; and (4) technologies easy to understand by the public. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, the Japan Atomic Power Company (JAPC) supported by the other Japanese PWR Utilities, Electricite de France (EdF), Westinghouse (WH) and Mitsubishi Heavy Industry (MHI) have studied application of passive technologies at a power rating of about 1,000 MWe. The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Using the AP-600 reference design as a basis, the authors enlarged the plant size to 3 -loops and added engineering features to conform with Japanese practice and Utilities' preference. The Simplified PWR (SPWR) program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1,000 MWe and 3 loops

  19. Fuel reliability of Bohunice NPP

    International Nuclear Information System (INIS)

    Paper summarizes experience from last 15 years of operation at NPP Jaslovske Bohunice. During this period, leaking fuel assemblies have had been identified by in-core sipping method and verified by vendor specified canister sipping method. Methodology of operational and outage fuel integrity monitoring is described. Full survey of identified leaking assemblies is given. Fuel failure rates are calculated separately for V-1 (V-230 type) and V-2 (V-213 type) units. Systematic difference - significantly lower fuel failure rate at V-213 units exists for all period investigated. Analysis of potential fuel failure reasons and all related measures (planned and already implemented) are presented. Design, operation and fabrication features have been analyzed with the aim to identify dominant factors contributing to fuel failure. No unambiguous reasons have been found so far. It is believed that there is a superposition of several factors and differences causing higher failure rate at V-230 type units. (author)

  20. KEPIC application on Korean NPP

    International Nuclear Information System (INIS)

    In the management of nuclear power plant, the security of safety is most important and the such a safety security is closely related to the governing code requirements for nuclear facility. In the first stage of NPP construction in Korea, there was no independent Korean codes for the nuclear facility, accordingly different kind of foreign codes were applied. From the later of 1980, KHNP leads the development of KEPIC (Korea Electric Power Industry Code). The development had been being performed in the three step and finished in the end of Dec. 2000. After that, the KEPIC developed had been selectively applied for the UI-Chin 5 and 6 units construction and it is now expected that the application of KEPIC will be markably expended in the Shin-Kori 1 and 2 and Shin Wol-Sung 1 and 2 units scheduled. Thereby here I introduce the status of development and application of the KEPIC for information of persons interested

  1. Does climate directly influence NPP globally?

    Science.gov (United States)

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

    2016-01-01

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

  2. Krsko NPP radioactive waste characteristics

    International Nuclear Information System (INIS)

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

  3. Introduction of Applying Technology for Decommissioning and Decontamination in PWR Npp's

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dukwon; Jo, Youngsoo; Kim, Seungil; Park, Jongsuk; Kim, Hyunki; Heo, Jun; Lee, Seban [Huviswater corporation, Siheung (Korea, Republic of)

    2015-05-15

    The technologies utilized in nuclear power plant decommissioning are classified into four categories such as decontamination, transaction, restoration and delivery division and the safety and economic feasibility of nuclear power plant decommissioning can be hugely affected depending on the selected technology. Therefore, this article aims to introduce new technologies that can be applied in the domestic nuclear power plant decommissioning service facility based on the dismantling and decontaminating technique used in Japan's WDF (Waste Dismantling Facility). As nuclear power plants are getting older, interests on a dismantling technique are increasingly attracting more attention. Decommissioning and decontamination business will be opened a new prospect in the field of nuclear industry in near future.

  4. Methodology on ageing management review for main components of a PWR NPP

    International Nuclear Information System (INIS)

    According to the requirements of NNSA, for Chinese operational NPPs periodical safety review (PSR) should be carried out every 10 years. Ageing management is one of the important safety factors to be reviewed. Entrusted by Qinshan Nuclear Power Plant (QNPP), Shanghai Nuclear Engineering Research and Design Institute (SNERDI) carried out the ageing management review (AMR), as a part of the first PSR of QNPP, from 2001 to 2003. This paper summarizes the methodology of the AMR process, including screening of critical components and structures, identification of main ageing mechanisms and their indicators and the tabulated review process, etc. 15 components and structures, hereafter referred as equipment, were selected as review objects based on their significance of safety, replaceability and cost-benefit considerations. To these objects, the main ageing mechanisms and relevant ageing indicators were identified according to specific working and environmental condition, design and manufacture information, operation and maintenance history, etc. The review can be divided into two parallel parts, the review for specific equipment and the review for overall management procedures and their implementation. To typical components, such as RPV and SG, fatigue analysis based on operational transient accounting was carried out to observe the actual safety margins. Through the review, the weaknesses in ageing management and potential threats to structural integrity were identified and thus continued improvement can be made in the next period of 10 years. (authors)

  5. Analysis of therma fatigue due to thermal stratification in a NPP steam generator injection nozzle

    International Nuclear Information System (INIS)

    This work is related to an experimental thermal stratification study aiming to quantify thermal fatigue damages in the pipe material. Thermal fatigue damages appear as a consequence of non-linear longitudinal and circumferential loads and thermal stripping present in pipes with thermal stratified flows. Thermal stratification phenomenon is present in pipelines of Nuclear Power Plants (NPP) and calculations done up to the years 80 just consider linear loads. Consequently, many NPP pipelines became failing. In this work an experimental section, simulating the injection nozzle of a NPP steam generator, is subjected to the effects of thermal fatigue due to thermal stratification. The experimental section is made of stainless steel pipe type AISI 304L and its geometric characteristics allowed the same range of Froude numbers of a Pressurized Water Reactor (PWR) NPP. Temperatures are measured externally and internally in three positions and deformations just externally in seven positions. Inside the pipe thermocouples are positioned vertically along the diameter in different levels. Deformations of the pipe experimental section are utilized as a guide parameter to carry out fatigue tests. Preliminary numerical simulations were done using a coupled analysis in the ANSYS code with temperatures and pressure inputs taken from thermo-hydraulic experimental results. The objectives in this work are quantify the thermal fatigue intensity imposed to the pipe material by thermal stratification experiments, verify the agreement between numerical and experimental thermal stratification results and obtain stresses and strain parameters to carry out fatigue tests in specimens made of pipe experimental section and in specimens made of the virgin pipe. In this work is possible to conclude that thermal stratification happens in the experimental section and that numerical and experimental results agreed in the pipe region where they are compared and that thermal stratification induces

  6. Optimizing NPP performance and service life

    International Nuclear Information System (INIS)

    The most effective way for new power production in Ukraine is the completions of the Khmelnitskij 2 and Rovno 4 NPP project. The report presents the financing terms and conditions of the Energoatom corporate bonds issue

  7. Krsko NPP Periodic Safety Review program

    International Nuclear Information System (INIS)

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

  8. Treatment of NPP wastes using vitrification

    International Nuclear Information System (INIS)

    Glass-based materials to immobilize various liquid and solid radioactive wastes generated at nuclear power plants (NPP) were designed. Glassy waste forms can be produced using electric melting including a cold crucible melting. Leach rate of cesium was found to be 10-5-10-6 g/(cm2 day) (IAEA technique). Volume reduction factor after vitrification reached 4-5. Various technologies for NPP waste vitrification were developed. Direct vitrification means feeding of source waste into the melter with formation of glassy waste form to be disposed. Joule heated ceramic melter, and cold crucible were tested. Process variables at treatment of Kursk, Chernobyl (RBMK), Kalinin, Novovoronezh (VVER) NPP wastes were determined. The most promising melter was found to be the cold crucible. Pilot plant based on the cold crucibles has been designed and constructed. Solid burnable NPP wastes are incinerated and slags are incorporated in glass. (author)

  9. Suomi NPP Ground System Performance

    Science.gov (United States)

    Grant, K. D.; Bergeron, C.

    2013-12-01

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

  10. Visaginas NPP Project Regional Approach: Lithuania

    International Nuclear Information System (INIS)

    Lithuania has a long standing nuclear energy history. The country is the host of the Ignalina NPP consisting of two RBMK-1500 reactors (a type of boiling water reactor developed by the Soviet Union) located in Visaginas, Lithuania. Ignalina NPP (INPP) Unit 1 came online in December 1983 and Unit 2 was completed in 1987. Lithuania agreed to close the Ignalina NPP as part of its Accession Treaty to the European Union of 2003, as the Ignalina NPP design shares similarities with the Chernobyl NPP. Unit 1 was closed in December 2004 and Unit 2 was closed on 31 December 2009. Around 80% of electricity production in Lithuania in 2009 came from Unit 2 of the INPP. However, following the closure of the Ignalina NPP, Lithuanian electricity net import was 62% of the entire electricity demand in 2010 and 59%32 in 2011. To meet its energy needs following the INPP’s closure, in the absence of a new nuclear power plant, Lithuania relies on a combination of imported electricity, predominantly from interconnections with the UPS/IPS network, and power from alternative domestic generation facilities, which are predominantly fossil plants reliant on gas or oil imports from other countries

  11. Practice for the upgrading of Trino Vercellese NPP: Technical and economical aspects

    International Nuclear Information System (INIS)

    In this report the experience gained in seismic re-evaluation of an old NPP (Trino Vercellese) is described. This PWR plant was not seismically designed. The main purpose of the upgrading, from the point of view of the Italian Directorate for Nuclear Safety - ENEA/DISP, was to have guaranteed the plant capability of achieving and maintaining a safe cold shutdown condition after a SSE seismic event. The main steps of the seismic review are discussed: definition of the new input motion; selection of structures, systems and components essential for a safe cold shutdown; definition of Codes and evaluation methods; seismic qualification of systems and components. Finally some modifications of a number of plant systems are described together with economical aspects. (author)

  12. Decommissioning Study of Oskarshamn NPP

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-06-15

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

  13. Decommissioning study of Forsmark NPP

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-06-15

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

  14. Decommissioning study of Forsmark NPP

    International Nuclear Information System (INIS)

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

  15. Decommissioning Study of Oskarshamn NPP

    International Nuclear Information System (INIS)

    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

  16. Nuclear disasters at Chornobyl NPP, Fukushima NPP and nuclear power engineering in the 21- century

    International Nuclear Information System (INIS)

    The article presents a brief analysis of nuclear accidents at the Chornobyl NPP 91986) and Fukushima NPP (2011), discusses causes and scenarios of the accidents. The radioactive contamination of the environment resulting from the disasters is characterized, and top-priority actions for mitigation of the consequences and protection of public are discussed

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

    Science.gov (United States)

    Doughty, C.

    2015-12-01

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

  18. Regulatory aspects of NPP safety

    International Nuclear Information System (INIS)

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

  19. Kozloduy NPP intranet portal, Bulgaria

    International Nuclear Information System (INIS)

    The Kozloduy NPP intranet portal was established in the late 1990s. The purpose of the portal was to provide access to general and frequently used information necessary for routine and daily work of the plant staff. Over the years the intranet site has been continuously improved and extended. The portal is now a standard tool for every member of plant personnel. The portal architecture has been designed on a modular basis and follows the general structure of the plant. The home page contains general, publicly accessible and frequently used information and corresponding links. Each major division maintains its own sub-portal, which services the specific needs of the division personnel. Hierarchical structure, pull down and shortcut menus facilitate navigation and provide a user friendly interface. The portal is based on FrameWork 1.1 and DotNetNuke and provides group and individual communications and data exchange. Most of the major plant databases related to documentation, plant operation, plant safety, plant systems data, training and human resources are accessible through the portal. Miscellaneous information and useful internal and external links also are available. Different types of communication services are organized through a separate server. Depending on their role and position, each staff member has been provided with an internal and/or external email address and an individually configured internet connection. For general purposes cable internet is accessible at several points, which are evenly located around the site, and there is also a secure wireless network connection. Search and retrieve functions are implemented through respective engines, which are incorporated into applications. The portal has a strongly defined access rights system. Anonymous access is prevented; page personalization is available only for limited specific cases. Figures show the home page and the path to Units 5 and 6 on-line technical parameters

  20. Integrated ageing management of Atucha NPP

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  1. Simulator experiments: effects of NPP operator experience on performance

    Energy Technology Data Exchange (ETDEWEB)

    Beare, A.N.; Gray, L.H.

    1984-01-01

    During the FY83 research, a simulator experiment was conducted at the control room simulator for a GE Boiling Water Reactor (BWR) NPP. The research subjects were licensed operators undergoing requalification training and shift technical advisors (STAs). This experiment was designed to investigate the effects of senior reactor operator (SRO) experience, operating crew augmentation with an STA and practice, as a crew, upon crew and individual operator performance, in response to anticipated plant transients. Sixteen two-man crews of licensed operators were employed in a 2 x 2 factorial design. The SROs leading the crews were split into high and low experience groups on the basis of their years of experience as an SRO. One half of the high- and low-SRO experience groups were assisted by an STA. The crews responded to four simulated plant casualties. A five-variable set of content-referenced performance measures was derived from task analyses of the procedurally correct responses to the four casualties. System parameters and control manipulations were recorded by the computer controlling the simulator. Data on communications and procedure use were obtained from analysis of videotapes of the exercises. Questionnaires were used to collect subject biographical information and data on subjective workload during each simulated casualty. For four of the five performance measures, no significant differences were found between groups led by high (25 to 114 months) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have significantly shorter task performance times than crews led by high experience SROs. The presence of the STA had no significant effect on overall team performance in responding to the four simulated casualties. The FY84 experiments are a partial replication and extension of the FY83 experiment, but with PWR operators and simulator.

  2. The Role of CVR in the Fuel Inspection at Temelin NPP

    International Nuclear Information System (INIS)

    Since first reload, NPP Temelin together with the fuel vendor (Westinghouse Electric Company LLC) is performing post-irradiation inspection on the fuel assemblies as additional proof of PWR material compatibility in VVER water chemistry. However, after ten years of successful operation the fuel vendor is changing and new plans for the fuel inspection are ready. Paper describes the past experiences with the fuel inspections and repairs at the NPP Temelin and the role of Research Centre Rez, Ltd. (CVR) in the cooperation with the new fuel vendor.In addition, Research Centre Rez Ltd. is a non-profit organization devoted to activities that require research reactors LR-0, LVR-15 and several experimental loops and devices. The Centre was founded in 2003 by the Czech Ministry of Education that fully endorsed a research proposal for all scientific and R and D activities. It is a unique institution providing a sophisticated infrastructure for research and development focused on advancement of analytical methods. The main purpose of the Research Reactors Division in CVR with aim of the fuel on-site post-irradiation inspection and water chemistry research are also presented in paper. (author)

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

    Institute of Scientific and Technical Information of China (English)

    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.

  4. Assessment and management of ageing of major nuclear power plant components important to safety: PWR pressure vessels. 2007 update

    International Nuclear Information System (INIS)

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that effective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) is one of the most important issues for plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. IAEA-TECDOC-1120 documented ageing assessment and management practices for pressurized water reactor (PWR) reactor pressure vessels (RPVs) that were current at the time of its finalization in 1997-1998. Safety significant operating events have occurred since the finalization of the TECDOC, e.g. primary water stress corrosion cracking (PWSCC) of Alloy 600 control rod drive mechanism (CRDM) penetrations and boric acid corrosion/wastage of RPV heads, which threatened the integrity of the RPV heads. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1120 should be updated by incorporating those new events and their countermeasures. The objective of this report is to update IAEA-TECDOC-1120 in order to provide current ageing management guidance for PWR RPVs to all involved in the operation and regulation of PWRs and thus to help ensure PWR RPV integrity in IAEA Member States throughout their entire service life

  5. Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals: 2007 update

    International Nuclear Information System (INIS)

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that effective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) is one of the most important issues for plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. IAEA-TECDOC-1119 documents ageing assessment and management practices for PWR Reactor Vessel Internals (RVIs) that were current at the time of its finalization in 1997-1998. Safety significant operating events have occurred since the finalization of the TECDOC, e.g. irradiation assisted stress corrosion cracking (IASCC) of baffle-former bolts, which threatened the integrity of the vessel internals. In addition, concern of fretting wear of control rod guide tubes has been raised in Japan. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1119 should be updated by incorporating those new events and their countermeasures. The objective of this report is to update relevant sections of the existing IAEA-TECDOC- 1119 in order to provide current ageing management guidance for PWR RVIs to all involved in the operation and regulation of PWRs and thus to help ensure PWR safety in IAEA Member States throughout their entire service life

  6. Development status of nuclear power in China and fundamental research progress on PWR primary water chemistry in China

    International Nuclear Information System (INIS)

    China's non-fossil fuels are expected to reach 20% in primary energy ratio by 2030. It is urgent for China to speed up the development of nuclear power to increase energy supply, reduce gas emissions and optimize resource allocation. Chinese government slowed down the approval of new nuclear power plant (NPP) projects after Fukushima accident in 2011. At the end of 2012, the State Council approved the nuclear safety program and adjusted long-term nuclear power development plan (2011-2020), the new NPP's projects have been restarted. In June 2015, there are 23 operating units in mainland in China with total installed capacity of about 21.386 GWe; another 26 units are under construction with total installed capacity of 28.5 GWe. The main type of reactors in operation and under construction in China is pressurized water reactor (PWR), including the first AP1000 NPPs in the world (units 1 in Sanmen) and China self-developed Hualong one NPPs (units 5 and 6 in Fuqing). Currently, China's nuclear power development is facing historic opportunities and also a series of challenges. One of the most important is the safety and economy of nuclear power. The optimization of primary water chemistry is one of the most effective ways to minimize radiation field, mitigate material degradation and maintain fuel performance in PWR NPPs, which is also a preferred path to achieve both safety and economy for operating NPPs. In recent years, an increased attention has been paid to fundamental research and engineering application of PWR primary water chemistry in China. The present talk mainly consists of four parts: (1) development status of China's nuclear power industry; (2) safety of nuclear power and operating water chemistry; (3) fundamental research progress on Zn-injected water chemistry in China; (4) summary and future. (author)

  7. Lifetime evaluation of Bohunice NPP components

    International Nuclear Information System (INIS)

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

  8. Activity transport models for PWR primary circuits

    International Nuclear Information System (INIS)

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR's. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.)

  9. Program of monitoring PWR fuel in Spain

    International Nuclear Information System (INIS)

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  10. Thermohydraulic calculation of WWER-type NPP

    International Nuclear Information System (INIS)

    Technique of thermohydraulic calculation of the WWER-type NPP in unsteady processes is described. Effective algorithm for solving hydrodynamics equations without regard for acoustic effects permitting to use enough large time integration step is given. Calculation of two-dimensional temperature fields in fuel element is considered. Method for calculating a pressurizer, steam generators and pumps is described as well

  11. Safety culture development at Daya Bay NPP

    International Nuclear Information System (INIS)

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

  12. Works on ageing management for Cernavoda NPP

    International Nuclear Information System (INIS)

    The document presents a general overview of the activities realized by CITON in the area of ageing management for Cernavoda NPP components. There are considered especially safety related components. The initial activity in this field started with general analyses of ageing phenomena implication on NPP safety and establishment of a general program for ageing management. During 1994-2000 period a series of research and development works included in 'Nuclear Safety Research and Development Program' financed by Ministry of Industry and Resources allowed the completion of important steps in ageing management: condition indicators definition, selection of systems and components necessary of ageing monitoring and evaluation, establishment of initial reference values for selected components monitoring, etc. Starting with 2001 year the ageing program includes evaluation of the service life for main of components (one system after the other approach), evaluation of present surveillance and data collection and proposal for improvement by computerized data acquisition and processing, extension of system, ageing evaluation for the whole plant critical components, both from safety and availability point of view. This program, with proper support and cooperation from the NPP owner, will allow first evaluation of the whole Cernavoda NPP - Unit 1 ageing, safety and availability implications and presentation of recommendations for operating conditions and maintenance optimization. The document emphasizes also the status of Cernavoda units, from ageing management point of view and the necessary actions to be adopted (in CITON opinion). (authors)

  13. RHR system reliability analysis of Krsko NPP

    International Nuclear Information System (INIS)

    In this paper Systems reliability analysis is applied to residual heat Removal System in Krsko NPP. Fault tree method is used. Qualitative analysis of the fault tree was made using FTAP-2 computer code, and quantitative using IMPORT code. results are evaluated and their possible application is given. (author)

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

    International Nuclear Information System (INIS)

    This annex deals with project control. Mochovce NPP suffered considerable delay primarily due to lack of money. This situation was corrected and construction resumed in 1996. Throughout the 'dormant' period the plant received considerable support from the major contractors, who maintained skeleton staff at site. Significant safety and managerial improvements are being introduced and a strategic plan for the plant has been developed. (author)

  15. Sitting Safety Aspects of Second Romanian NPP

    International Nuclear Information System (INIS)

    The first Romanian NPP CANDU 6 type reactor gone to erection in 1980 on Cernavoda site planned to have 5 units like the Wolsong applied design project for nuclear island. For the BOP parts the ASALDO-GE project was applied with the careful about the interface connection NSP requirements. The new NPP sitting studies began from 1982 in a serious manner as first part on Nuclear Power Plant Romanian Program adopted by political and governmental authorities at the time. For develop the all package of the studies in concordance with the first IAEA Safety Standards recommendations. Till the 1982 the first mission of design and research multi-branch of specialists team was to adapt the NPP Cernavoda project having a open water cooling circuit to the new parameters of close water cooling circuit. But the team was looking at the other type of NPP for sitting. Also in the same time was studied the possibility of NSP foundation on hard less or soft soil foundation strata in connection with safety aspects. The close circuit of cooling water means others parameters of systems and need very large cooling towers. Also must be reconsidering the safety systems design and performance as new solution. In the south of Transylvania historical region in Romania the Olt River run from west to east having medium multi annual flow around 70 m3/s. The Olt River has a chain of small hydropower in operation and other planned. From geological and geophysical points of view two main faults, along the Olt river valley, one of this having seismically small activities was detected. Site region geotechnical studies show small quantity underground natural gas, salt and peat. The initial nuclear program has imposed 4 NPP units site near Olt River. Taking into account the orogenesis, water cooling needs and other local feature can't be built more than two NPP units on a site. This paper tries to reconsider the old analysis from the last IAEA Safety Standards point of view taking into account the new

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

    Lifescience Database Archive (English)

    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 ...

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

    Lifescience Database Archive (English)

    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 ...

  18. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  19. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.)

  20. A historical survey of the Ignalina NPP

    International Nuclear Information System (INIS)

    Full text: When the boom of nuclear power industry began in the former Soviet Union, the idea of constructing the Ignalina NPP occurred to the circles in Moscow's central institutions at the turn of the 1970s. The nuclear power plant remained a facility under all-union jurisdiction supervised by the Ministries of Atomic Energy and Medium-Machine Building of the USSR from September 16, 1971, when the Central Committee of the Communist Party of the Soviet Union and the Council of Ministers of the Soviet Union adopted the resolution regarding the beginning of its construction, until Lithuania regained independence in 1990. Nuclear power is the basis of Lithuania's power industry. The Ignalina NPP is a product of the former Soviet Union. Two reactors of RBMK-1SOO type are operational at the Ignalina NPP. This is the most advanced and the most recent version of the RBMK reactor design series (only two reactors of this type have ever been built). The power plant was built as part of the Soviet Union's North-West Unified Power System rather than to meet Lithuania's needs. The first unit of Ignalina NPP was commissioned in late 1983, and the second one in August 1987. A total of four units with RBI/1K-1S00 reactors were to be built. However, due to political and safety motives the construction of the third unit was suspended as early as 1989. After Lithuania declared independence in 1990, the Ignalina NPP was still guarded by Soviet troops and KGB operatives, and remained under the jurisdiction of the Soviet Union until August 1991. Supervision was carried out by that country's regulatory authority, the State Nuclear Power Supervision Inspection (Gosatomnadzor). It was only after the political events of August 1991 in Moscow that the Ignalina NPP finally came under the authority of the Lithuanian Republic. It is now controlled administratively by the Lithuanian Ministry of Economy, and its supervision is carried out by the Lithuanian State Nuclear Power Safety Inspectorate

  1. Adding a much needed 300 MWe at South Africa's Arnot coal fired power plant

    Energy Technology Data Exchange (ETDEWEB)

    Rich, G. [Alstom, Rugby (United Kingdom)

    2008-12-15

    As power stations built in the last thirty years approach the end of their design life, and the cost of new capacity continues to increase, along with demands for improved efficiency and lower emissions, an integrated approach to retrofit looks increasingly compelling. The ambitious upgrade project currently underway at the Arnot coal fired plant in South Africa, which will result in an update from 6 x 350 MWe to 6 x 400 MWe and a life extension of 20 years, illustrates the benefits. 2 figs.

  2. Method and Result of Experiment for Support of Technical Solutions in the Field of Perfection of a Nuclear Fuel Cycle for Future PWR Reactors

    International Nuclear Information System (INIS)

    The paper presents the basics of approach of planning and carrying out of experiments to validate safety PWR reactors of the future when accepting technical solutions concerning using of improved fuel rods in fuel assembly. Basic principles and criteria used for the validation of technical solutions and developments in improving of nuclear fuel cycle of PWR reactors of the future are presented from the point of safety of future operation of modified fuel rods. We explore the questions of safety operation of PWR reactors with fuel assemblies, containing fuel rods with different length of fuel. The paper discusses the ways of solving of important tasks of critical facility experiments conducting for verification of new technical solutions in the sphere of PWR nuclear fuel cycle improvement on the base of international standards ISO 2000:9000 and functional safety recommendations of IEC (International Electromechanical Commission). New Federal laws of Russian Federation define the main principle for demands to NPP and any supplier of nuclear techniques. The principle is 'quantity indicators of risk should not exceed comprehensible social size of the established indicators of safety for any moment of operation of NPP'. On the other hand the second principle should be applied to extraction of the greatest benefit from operation of the equipment, systems or the NPP as whole: 'The long operation and full commercial use of resource and service properties of the equipment, systems and the NPP as a whole'. Realization of this principle assumes development and introduction of new technical solutions for a validation of guarantees of safety of the future operation of NPP or it separate components. Solving the practical problems of a validation of safety use of fuel rods with the increased length of a fuel column in fuel assembly in nuclear reactors of the future, we should choose new strategies and programs of verification experiments on the base of the analysis of guarantees

  3. NPP bulk equipment dismantling problems and experience

    International Nuclear Information System (INIS)

    NPP bulk equipment dismantling problems and experience are summarized. 'ECOMET-S' JSC is shown as one of the companies which are able to make NPPs industrial sites free from stored bulk equipment with its further utilization. 'ECOMET-S' JSC is the Russian Federation sole specialized metallic LLW (MLLW) treatment and utilization facility. Company's main objectives are waste predisposal volume reduction and treatment for the unrestricted release as a scrap. Leningrad NPP decommissioned main pumps and moisture separators/steam super heaters dismantling results are presented. Prospective fragmentation technologies (diamond and electro-erosive cutting) testing results are described. The electro-erosive cutting machine designed by 'ECOMET-S' JSC is presented. The fragmentation technologies implementation plans for nuclear industry are presented too. (author)

  4. Review of site data for Cernavoda NPP

    International Nuclear Information System (INIS)

    The review of CANDU-PHWR-700 standard design according to the geological and seismic conditions of Cernavoda NPP site was made during the period 1980-1985 based on the geological, geotechnical, seismological design data as well as based on the design data for ground dynamics assessed in 1980. During the construction period of Unit 1 and partially of Unit 2-5, some observations ad in-situ and lab determinations have been done, resulting in additional information as to the ones determined during the period of site studies. Considering all the geological, geotechnical, seismic and ground dynamics data obtained both during the site investigation period and the construction period of Unit 1 and partially of Unit 2-5, a review of the site data was made in order to determine a representative set of data for Cernavoda NPP site, as well as to reduce the conservatory degree existing in the initial data. (author). 1 fig., 1 tab., 6 refs

  5. Intelligent system for accident identification in NPP

    International Nuclear Information System (INIS)

    Accidental situations in NPP are great concern for operators, the facility, regulatory bodies and the environmental. This work proposes a design of intelligent system aimed to assist the operator in the process of decision making initiator events with higher relative contribution to the reactor core damage occur. The intelligent System uses the results of the pre-operational Probabilistic safety Assessment and the Thermal hydraulic Safety Analysis of the NPP Juragua as source for building its knowledge base. The nucleus of the system is presented as a design of an intelligent hybrid from the combination of the artificial intelligence techniques fuzzy logic and artificial neural networks. The system works with variables from the process of the first circuit, second circuit and the containment and it is presented as a model for the integration of safety analyses in the process of decision making by the operator when tackling with accidental situations

  6. SAT for NPP personnel training in USA

    International Nuclear Information System (INIS)

    This discussion addressed the experience with the application of SAT at USA NPPs. In particular, the transition of NPP training processes, staff composition, and reporting structure from the TMI accident to present. As well, oversight and guidance activities of the INPO and more intensive inspection by the NRC began during this period. The average NPP training staff grew to 30-40 per unit, along with a change in reporting line from plant to corporate management. With the reduction of resources occurring in the late 1980s, overall training staff size decreased, the composition changed, and reporting line reverted to plant management. The overall lessons-learned for application of the SAT consisted of the need for simplification, management involvement, and exploitation of the technology

  7. LTO License Application Project NPP Borssele

    International Nuclear Information System (INIS)

    Borssele NPP plans to extend its operating life with 20 years until 2034. Borssele has started the project LTO 'bewijsvoering' (LTO 'Justification') in order to meet the requirements of the Dutch regulator. The outline of the project is based on IAEA safety guide 57 'Safe Long Term Operation of Nuclear Power Plants'. This paper describes the contents and coherence of the different parts in the project and how these respond to the IAEA guidelines on LTO. The goal of the project LTO 'bewijsvoering' is to ensure that safety and safety relevant systems, structures and components continue to perform their intended functions during long term operation. The outcome of the project LTO 'bewijsvoering' will be used for a license change application and this will be submitted to the Dutch regulator KFD for approval of prolonged operation of Borssele NPP after 2013. (author)

  8. NPP EQUIPMENT LIFETIME MANAGEMENT UNDER AGEING

    Directory of Open Access Journals (Sweden)

    Olga M. Gulina

    2012-01-01

    Full Text Available This paper deals with the problem of nuclear power plant (NPP equipment lifetimeprediction with the use of information about damage processes, operation conditions andpreventive actions. The developed model is based on Kalman linear stochastic filter. For steamgenerator heat-exchanged tubes we use the martingale theory for predicting the number ofsuppressed heat-exchanged tubes to the next control. Finally, we formulate the principle of SGlifetime optimal management.

  9. Decommissioning database of V1 NPP

    International Nuclear Information System (INIS)

    Since 2001, the preparation of V1 NPP practical decommissioning has been supported and partly financed by the Bohunice International Decommissioning Support Fund (BIDSF), under the administration of the European Bank for Reconstruction and Development. AMEC Nuclear Slovakia, together with partners STM Power and EWN GmbH, have been carrying out BIDSF B6.4 project - Decommissioning database development (June 2008 until July 2010). The main purpose of the B6.4 project is to develop a comprehensive physical and radiological inventory database to support RAW management development of the decommissioning studies and decommissioning project of Bohunice V1 NPP. AMEC Nuclear Slovakia was responsible mainly for DDB design, planning documents and physical and radiological characterization including sampling and analyses of the plant controlled area. After finalization of all activities DDB includes over 75.000 records related to individual equipment and civil structures described by almost 3.000.000 parameters. On the basis of successful completion of the original contract the amendment was signed between JAVYS and Consultant's Consortium related to experimental characterization of NPP activated components. The works within this amendment have been still running. (authors)

  10. Knowledge management in the NPP domain

    Energy Technology Data Exchange (ETDEWEB)

    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)

  11. Tritium liquid effluents from the Krsko NPP

    International Nuclear Information System (INIS)

    In the past, 12-months' fuel cycles in the Krsko NPP had not caused any problems regarding compliance with its Technical Specifications and license limits on liquid tritium releases (20 TBq/year, 8 TBq/three months). The first 18-months' fuel cycle, which was introduced in 2004, required fuel with higher enrichment, higher boron concentration in the primary coolant and more fuel rods with burnable poisons. In 2005, the NPP operated without refueling outage for the whole year and produced the highest amount of energy so far. Due to these facts and a few unplanned shutdowns and power reductions, production of tritium and releases increased strongly in 2005. As a result, the Krsko NPP hardly succeeded to stay within regulatory limits on tritium releases. However, the three-months' limit was exceeded in the first quarter of 2006. On the basis of conclusions acquired from the SNSA's study and practice of other European countries the SNSA considerably increased the annual limit of permitted liquid tritium releases (from 20 TBq to 45 TBq) and abolished the three-months' limit. At the same time, the SNSA reduced the limit of fission and activation products by halves. (author)

  12. Knowledge management in the NPP domain

    International Nuclear Information System (INIS)

    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)

  13. 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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Institute of Scientific and Technical Information of China (English)

    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.

  15. Safety Culture Survey in Krsko NPP

    International Nuclear Information System (INIS)

    The high level of nuclear safety, stability and competitiveness of electricity production, and public acceptability are the main objectives of Krsko Nuclear Power Plant. This is achievable only in environment where strong Safety Culture is taking dominant place in the way how employees communicate, perform tasks, share their ideas and attitudes, and demonstrate their concern in all aspects of work and coexistence. To achieve these objectives, behaviour of all employees as well as specific ethical values must become more transparent and that must arise from the heart of organization. Continuous ongoing and periodic self assessments of Safety Culture in Krsko NPP present major tools in implementation process of this approach. Benefits from Periodic interdisciplinary focused self assessment approach, which main intention is finding the strengths and potential areas for improvements, was used second time to assess the area of Safety Culture in Krsko NPP. Main objectives of self assessment, performed in 2006, were to increase the awareness of the present culture, to serve as a basis for improvement and to keep track of the effects of change or improvement over a longer period of time. For the purpose of effective self assessment, extensive questionnaire was used to obtain information that is representative for whole organization. Wide range of questions was chosen to cover five major characteristics of safety culture: Accountability for safety is clear, Safety is integrated into all activities, Safety culture is learning-driven, Leadership for safety is clear and Safety is a clearly recognized value. 484 Krsko NPP employees and 96 contractors were participated in survey. 70-question survey provided information that was quantified and results compared between groups. Anonymity of participant, as well as their willingness to contribute in this assessment implicates the high level of their openness in answering the questions. High number of participant made analysis of

  16. Thorium fuel cycle study for PWR applications

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Jae Yong; Kim, Myung Hyun [Kyung Hee Univ., Seoul (Korea, Republic of)

    1997-12-31

    A nuclear design feasibility of thorium fueled high converting PWR was investigated. Two kinds of fuel design option were tested for the comparison with conventional UO{sub 2} fuel. The first one was an application of MHTGR pyro-carbon coated particle fuels. The other design was an application of MOX fuels as a ThO{sub 2}-PuO{sub 2} ceramic pellet. In the case of carbon-coated particle fuels, there was no benefit in nuclear design aspect because enrichment of U-235 was required over 5 w/o in order to match with the K-infinite of Ulchin-3/4 fuels. However, the use of thorium based plutonium fuels in PWR gave favorable aspects in nuclear design such as flatter K-infinite curve, lower M. T. C. and lower F. T. C. than that of UO{sub 2} fuel. (author). 6 refs., 3 tabs., 6 figs.

  17. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  18. The integrated PWR; Les REP integres

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  19. Shielding design for PWR in France

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    The activation and transport of corrosion products around a PWR circuit is a major concern to PWR plant operators as these may give rise to high personnel doses. The understanding of what controls dose rates on ex-core surfaces and shutdown releases has improved over the years but still several questions remain unanswered. For example the relative importance of particle and soluble deposition in the core to activity levels in the plant is not clear. Wide plant to plant and cycle to cycle variations are noted with no apparent explanations why such variations are observed. Over the past few years this group have been developing models to simulate corrosion product transport around a PWR circuit. These models form the basis for the latest version of the BOA code and simulate the movement of Fe and Ni around the primary circuit. Part of this development is to include the activation and subsequent transport of radioactive species around the circuit and this paper describes some initial modelling work in this area. A simple model of activation, release and deposition is described and then applied to explain the plant behaviour at Sizewell B and Vandellos II. This model accounts for activation in the core, soluble and particulate activity movement around the circuit and for activity capture ex-core on both the inner and outer oxides. The model gives a reasonable comparison with plant observations and highlights what controls activity transport in these plants and importantly what factors can be ignored. (authors)

  1. Corporate portal system at PAKS NPP, Hungary

    International Nuclear Information System (INIS)

    The new Corporate Portal System (CPS) of Paks NPP was launched in November 2006. The portal is based on one of the latest technologies, Plumtree Enterprise WEB 5.0. The main purpose of the installation of the new technology was to serve the working culture change, to give a platform to access all information and applications including the integrated process model used at the NPP. The new technology also supports those goals which were defined in the organization development programme: e.g. to improve internal communication with the establishment of communities of practice. Installation of the CPS has provided a powerful tool for knowledge management; it is possible to share and find all information through a controlled access in documents from various sources, to have links to people, portlets and different communities. Document management of the Paks NPP is supported by the integration of the Document 5 application, as the new Electronic Data Management System (EDMS) and the CPS. Depending on their access rights, all users of the CPS, through Microsoft Internet Explorer, can access technical, economic and human resources documents which are stored anywhere on the internal network (file servers, EDMS, old INRANET). The CPS is also accessible from the internet through a secure connection. The main concept is the integration of all applications to one platform and to help users to find all information they need. An access control list specifies which users and groups have access to an object (and what kind of access privileges they have such as read, select, edit, admin)

  2. Mochovce NPP safety improvement and completion

    International Nuclear Information System (INIS)

    6th Nuclear society information meeting dealt with the completion of the Mochovce NPP with regard to implementation of safety measures. It was aimed to next problems: I. 'Survey' presentation on the situation of the nuclear power industry in partner countries; II. Basic technical presentations; III. Presentations of operators of the other VVER 440/213 NPPs on their activities in the field of safety improvement in relation to IAEA recommendations; IV. Technical solutions of safety improvements ranked with IAEA degree 3 (Report SC 108 VVER); V: Technical solutions of selected Safety Measures ranked with IAEA degree 2 and 1 (Report SC 108 VVER)

  3. Nuclear Oversight Function at Krsko NPP

    International Nuclear Information System (INIS)

    The nuclear oversight function is used at the Krsko NPP constructively to strengthen safety and improve performance. Nuclear safety is kept under constant examination through a variety of monitoring techniques and activities, some of which provide an independent review. The nuclear oversight function at the Krsko NPP is accomplished by Quality and Nuclear Oversight Division (SKV). SKV has completed its mission through a combination of compliance, performance and effectiveness-based assessments. The performance-based assessment is an assessment using various techniques (observations, interviews, walk-downs, document reviews) to assure compliance with standards and regulations, obtain insight into performance, performance trends and also to identify opportunities to improve effectiveness of implementation. Generally, the performance-based approach to oversight function is based on some essential elements. The most important one which is developed and implemented is an oversight program (procedure). The program focuses on techniques, activities and objectives commensurate with their significance to plant operational safety. These techniques and activities are: self-assessments, assessments, audits, performance indicators, monitoring of corrective action program (CAP), industry independent reviews (such as IAEA's OSART and WANO Peer Review), industry benchmarking etc. Graded approach is an inherent product of a performance based program and ranking process. It is important not only to focus on the highest ranked performance based attributes but to lead to effective utilization of an oversight program. The attributes selected for oversight need to be based on plant specific experience, current industry operating experience, supplier's performance and quality issues. Collaboration within the industry and effective utility oversight of processes and design activities are essential for achieving good plant performance. So the oversight program must integrate relevant

  4. Safe 15 Terawatt of Temelin NPP

    International Nuclear Information System (INIS)

    In this work author presents a project Safe 15 Terawatt realised on the Temelin NPP. This project is one of the eight key projects of the CEZ group, associated in the 'Programme of efficiency'. The project started in June 2007 with long-term goals for horizon of year 2012. The safety indicators will be reached of the first quarter level of world's nuclear power plant - by the end of the first decade. By the end of year 2012 we will have achieved annual production of 15 billion kWh - in the Czech Republic: 15 Terawatt.

  5. Sizewell: proposed site for Britain's first PWR power station

    International Nuclear Information System (INIS)

    The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

  6. Sociological investigations on Ignalina NPP and within its surroundings

    International Nuclear Information System (INIS)

    The purpose of this study was to determine the impact of Ignalina NPP and Visaginas town on the social territorial processes in the region and to reveal the impact of Ignalina NPP on the regional economic, social, demographic, political and cultural processes in the context of ecological and psychological affect. According to the results of this research three quarters of the inhabitants and the functionaries of local administration hold an opinion that operation of Ignalina NPP posses threat for the population and environment. Meanwhile they are sure that danger of Ignalina NPP is not critical. 21 - 35 % of the local administrators speak for the closure of Ignalina NPP , whereas half of Visaginas residents and three quarters of the local administrators indicate that operation of reactors is expedient. Over 90% of the population do not have sufficient information on the operation of Ignalina NPP. In the opinion of the rest Lithuanian people Ignalina NPP zone is related with the physical danger and the image of Visaginas residents as the 'others', 'strangers'. More than 90% of Ignalina NPP employees are Russian speaking, not native Lithuanians. The social relations of Visaginas with the environment are poor as a result of the situation of the town, lack of communications and cultural self isolation. (author)

  7. Trends and Variability of AVHRR-Derived NPP in India

    Directory of Open Access Journals (Sweden)

    Hirofumi Hashimoto

    2013-02-01

    Full Text Available In this paper, we estimate the trends and variability in Advanced Very High Resolution Radiometer (AVHRR-derived terrestrial net primary productivity (NPP over India for the period 1982–2006. We find an increasing trend of 3.9% per decade (r = 0.78, R2 = 0.61 during the analysis period. A multivariate linear regression of NPP with temperature, precipitation, atmospheric CO2 concentration, soil water and surface solar radiation (r = 0.80, R2 = 0.65 indicates that the increasing trend is partly driven by increasing atmospheric CO2 concentration and the consequent CO2 fertilization of the ecosystems. However, human interventions may have also played a key role in the NPP increase: non-forest NPP growth is largely driven by increases in irrigated area and fertilizer use, while forest NPP is influenced by plantation and forest conservation programs. A similar multivariate regression of interannual NPP anomalies with temperature, precipitation, soil water, solar radiation and CO2 anomalies suggests that the interannual variability in NPP is primarily driven by precipitation and temperature variability. Mean seasonal NPP is largest during post-monsoon and lowest during the pre-monsoon period, thereby indicating the importance of soil moisture for vegetation productivity.

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

    Institute of Scientific and Technical Information of China (English)

    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.

  9. NPP electrical price and tariff in the world

    International Nuclear Information System (INIS)

    Construction of a Nuclear Power Plant (NPP) is always become a controversial issue. Nuclear utility and other party which support the NPP present a calculation of NPP electricity cost too optimistic. However for utility and other party that contra to nuclear present a calculation of NPP electricity cost too pessimistic. This study present to reduce the controversy of nuclear cost. In this study, capital cost (Engineering Procurement Construction, EPC) was taken from Asian, America and Europe, operating and maintenance cost uses experience data of PLN, and nuclear fuel cost uses data year of 2008 with high price, low price and average price scenario. The methodological tools used to compare electricity generation cost was LEGECOST, a program developed by IAEA (International Atomic Energy Agency), while for electricity tariff- price calculation using a program developed by PLN research and development center. With the discount rate 10%, the result shows that the cheapest electricity generation cost of NPP is less than 40 mills/kWh, and average electricity tariff was 55 mills/kWh. In the Europe countries the electricity tariff more expensive than NPP in Asia. However generating cost and electricity tariff of NPP in United Stated of America (USA) less competitive because investment cost more expensive. Generating cost and electricity tariff was different at each country depend on salary, labor wage, materials price, construction specification, regulation related to NPP and environment aspect. (author)

  10. RCM at NPP Dukovany, Czech Republic

    International Nuclear Information System (INIS)

    The Project components are: RCM course for beneficiaries of the Project; software - EPRI RCM Workstation 2.5; pilot RCM projects at NPP Dukovany and NPP Bohunice. For the pilot project at Dukovany Failure Modes and Effects Analysis (FMEA) is performed. As a result a proposal of the current PM programme revision was worked up in a form of a comparison between the recent PM tasks and the RCM recommendations. The plant maintenance history electronic databases are found to be comprehensive, but not ready for the reliability determination. The maintenance history doesn't contain an identification of specific failure modes. Therefore, generic reliability data were used at FMEA (Failure Modes: Component Type, Failure Mode ID, Failure Mode Description, Probability Value, and Probability Type). In some important cases, when the RCM recommendation was to change a current task interval, the RCM proposed task frequencies were calculated from real plant data which were verified and consulted. Further, condensate pumps maintenance strategy and failure rate review is made. RCM analysis and recommendations for the pump maintenance are presented. The RCM application to selected electric systems is also discussed. Based on the calculations, maintenance programme was modified

  11. Utilization of NPP Krsko plant specific simulator

    International Nuclear Information System (INIS)

    NPP Krsko started with licensed operator training using its own plant-specific full scope simulator in April 2000. Today, two years after simulator was completed, the benefits of simulator use are visible in various fields. The simulator was effectively used to conduct licensed operator continuing training and practical examinations. Two-year continuous training program was designed to help maintain and improve operator performance. The simulator was also used to provide just-in-time training prior to plant evolutions. Together with licensed operators the non-licensed operators are also included into simulator training to provide affective team training opportunity and to foster good communication and increase scenario realism. Now, the first group of initial licensed operator training using plant-specific simulator is also almost completed. It is the first time that NPP Krsko training department conducted complete initial training and this will represent the great experience for future training. Besides training, the simulator was also utilized for procedure development and validation, operating standards development, testing of plant modifications and other activities, like emergency preparedness procedures validation and training exercises.(author)

  12. Project No. 10 - Partial restoration of Ignalina NPP territory

    International Nuclear Information System (INIS)

    At present Ignalina NPP territory makes a total of 2544 ha of land. Due to termination of construction activity development and due to the decision taken to shutdown unit 1 the need in such a territory fell off. For normal and safe operation of Ignalina NPP 1440 ha is enough, including 1237 ha for of Ignalina NPP administrative area and 203 ha for auxiliary objects. Ignalina NPP will have to rearrange territory, forestry that was damaged during the construction activities of the plant and to restore the damaged farmlands and to pass the rearranged forestry that belonged to the Ignalina NPP to the Ministry of Forestry. The total estimated cost of the project is about 1.042 M EURO

  13. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  14. Industrywide survey of PWR organics. Final report

    International Nuclear Information System (INIS)

    Thirteen Pressurized Water reactor (PWR) secondary cycles were sampled for organic acids, total organic carbon, and inorganic anions. The distribution and removal of organics in a makeup water treatment system were investigted at an additional plant. TOC analyses were used for the analysis of makeup water systems; anion ion chromatography and ion exclusion chromatography were used for the analysis of secondary water systems. Additional information on plant operation and water chemistry was collected in a survey. The analytical and survey data were compared and correlations made

  15. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

  16. Assessment and Management of ageing of major nuclear power plant components important to safety: PWR pressure vessels

    International Nuclear Information System (INIS)

    ageing management and economic planning. The target audience of the reports consists of technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. The NPP component addressed in the present publication is the PWR pressure vessel

  17. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing keff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR keff markedly. The PWR keff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  18. ATUCHA I NPP - Emergency drill practice

    International Nuclear Information System (INIS)

    Full text: Atucha I NPP performs an Emergency Drill Practice once a year. Its main goals are: -) Fulfill the requirements of the Argentine Nuclear Regulatory Authority (ARN) regarding Atucha I NPP's Operating License; -) Fulfill the commitment with the community regarding the safe and reliable operation Atucha I NPP; -) Verify the response of the Civil Organizations, Security Forces, and Armed Forces, as well as the correct application of the Emergency Plan; -) Perform the 'General Alarm Drill' periodic control; -) Perform a re-training of the members of the Security Advisor Internal Committee (CIAS) on the Internal and External Aspects of the Emergency Plan and on the related procedures; -) Test the Emergency Communications System. New goals are added every year, considering the Drill's scope. This drill comprises two different kinds of practices: Internal practices (practices in the station, with our personnel) and external practices (practices outside the station with governmental organizations). Internal practices comprise: -) Internal and external communications practices; -) Acoustic alarms; -) Personnel gathering in the Meeting Points; -) Safety of selected Meeting Points; -) Personnel count, selective evacuation; -) Iodide Potassium pills distribution; -) CICE (Internal Group for Emergency Control) Coordination. External practices comprise: -) Nuclear Regulatory Authority; -) Argentine Navy, Comando Area Naval Fluvial, Base Naval Zarate; -) Lima firemen; -) Zarate firemen; -) Municipal Civil Defense (Zarate and Lima); -) National Guard, Escuadron Atucha; -) Zarate Regional Hospital; -) Lima Police Department; -) Zarate Police Department; -) Argentine Coast Guard, Zarate; -) Local radios: Radio FM Libre, FM El Sitio; -) First Aid clinic. The following activities are performed together with the aforementioned organizations: -) Formation of an 'Operative committee'; -) Evacuation of citizens in a 3 km radio; -) Control of every access to Lima; -) Control of

  19. Thermal fatigue damage evaluation of a PWR NPP steam generator injection nozzle model subjected to thermal stratification phenomenon

    Energy Technology Data Exchange (ETDEWEB)

    Leite da Silva, Luiz, E-mail: silvall@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear, CDTN/CNEN, Av. Presidente Antonio Carlos, 6627, Campus UFMG, Pampulha Belo Horizonte, MG CEP 31.270-901 (Brazil); Rodrigues Mansur, Tanius, E-mail: tanius@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear, CDTN/CNEN, Av. Presidente Antonio Carlos, 6627, Campus UFMG, Pampulha Belo Horizonte, MG CEP 31.270-901 (Brazil); Cimini Junior, Carlos Alberto, E-mail: cimini@demec.ufmg.b [Departamento de Engenharia Mecanica da Universidade Federal de Minas Gerais, DEMEC/UFMG, Av. Presidente Antonio Carlos, 6627, Pampulha, Belo Horizonte, MG CEP 31.270-901 (Brazil)

    2011-03-15

    Thermal stratification phenomenon with the same thermodynamic steam generator (SG) injection nozzle parameters was simulated. After 41 experiments, the experimental section was dismantled; cut and specimens were made of its material. Other specimens were made of the preserved pipe material. By comparing their fatigue tests results, the pipe material damage was evaluated. The water temperature layers and also the outside pipe wall temperatures were measured at the same level. Strains outside the pipe in 7 positions were measured. The experimental section develops thermal stratified flows, stresses and strains caused enlargement of material grain size and reduction in fatigue life.

  20. Industry Operating Experience Process at Krsko NPP

    International Nuclear Information System (INIS)

    Experience has shown that number of minor events and near misses, usually without immediate or significant impact to plant safety and reliability, are precursors of significant or severe events due to the same or similar root or apparent cause(s). It is therefore desirable to identify and analyze weaknesses of the precursor problems (events) in order to prevent occurrence of significant events. Theoretically, significant events could be prevented from occurring if the root cause(s) of these precursor problems could be identified and eliminated. The Operating Experience Program identifies such event precursors and by reporting them to the industry, plant specific corrective actions can be taken to prevent events at other operational plants. The intent of the Operating Experience Program is therefore to improve nuclear power plant safety and reliability of the operating nuclear power plants. Each plant develops its own Operating Experience Program in order to learn from the in-house operating experience as well as from the world community of nuclear plants. The effective use of operating experience includes analyzing both plant and industry events in order to identify fundamental weaknesses and then determining appropriate plant-specific actions that will minimize the likelihood of similar events. Learning and applying the lessons from operating experience is an integral part of station safety culture and is encouraged by managers throughout the top plant administrative programs and procedures. Krsko NPP is developed it own Operating Experience Program by using the most relevant INPO/WANO/IAEA guidelines as well as its own knowledge, skills an operating practice. The Operating Experience Program is a part of the Corrective Action Program, which is among top management programs, thus program is strongly encouraged by top management. The purpose of Operating Experience Program is to provide guidance for using, sharing, and evaluating operating experience information

  1. Students education and training for Slovak NPP

    International Nuclear Information System (INIS)

    Slovak University of Technology is the largest and also the oldest university of technology in Slovakia. It is certain that more than 50% of the highly-educated technicians who are currently working in the nuclear industry have graduated from this university. The Department of Nuclear Physics and Technology of the Faculty of Electrical Engineering and Information Technology as one of the seven faculties of this University feels the responsibility to impart proper engineering education and training for Slovak NPP operating staff. The education process is realised via undergraduate (BSc), graduate (MSc) and postgraduate (PhD) study as well as via specialised training courses within the framework of a continuous education system. (author)

  2. N-16 monitors: Almaraz NPP experience

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Basic information about development and construction of a PWR

    International Nuclear Information System (INIS)

    1.0) Plant layout of a PWR; 2.0) principle design of a PWR and the reactor coolant system; 3.0) reactor auxiliary and ancillary systems; 3.1) volume control system; 3.2) boric acid control and chemical feeding system; 3.3) coolant purification and degassing system; 3.4) coolant storage and treatment system; 3.5) nuclear component cooling system; 3.6) liquid waste processing system; 3.7) gaseous waste processing system; 4.0) residual heat removal system; 5.0) emergency feedwater system; 6.0) containment design; 7.0) fuel handling, storage and transport system in a PWR. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Use of interview and inquiry procedures in design process for NPP control room upgrade

    International Nuclear Information System (INIS)

    Major part of design activities for upgrade of NPP control rooms is using NPP personnel operating experience and participation of NPP personnel in detecting and solving the human factor problems. The ways of such participation include application of various human factors engineering and psychological techniques, for example interviews, inquiries (filling in questionnaire), operator activity observations. The present paper deals with activation of NPP personnel role in the design of control room upgrade, in particular, in creation of operator support systems. NPP personnel selection for incorporation in NPP upgrade group and initial training design skills are considered. NPP personnel responsibilities during particular stages of control room designing are specified. (author). 1 ref

  6. NPP safety in Slovakia according to stress tests after accident in Fukushimi

    OpenAIRE

    Králik, Juraj

    2013-01-01

    This paper presents the new requirements to test of the safety and reliability of the NPP structures due to the last nuclear accidents in the world. The accidents of the NPP in Chernobyl and Fukushima give us the new inspiration to verify the safety level of the NPP structures. The probabilistic assessment of NPP structures for PSA level 2 of VVER 440 in the case of LOCA accident is presented. The results of the probabilistic nonlinear analysis of the NPP structures are present...

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

    Institute of Scientific and Technical Information of China (English)

    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.

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

    CERN Document Server

    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...

  9. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

    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)

  10. Zebra: An advanced PWR lattice code

    International Nuclear Information System (INIS)

    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)

  11. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    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)

  12. Crevice chemistry control in PWR steam generators

    International Nuclear Information System (INIS)

    To establish a basis for predicting and eventually controlling crevice solution chemistry in PWR steam generators, hideout tests were performed at several units. Results indicated that impurity hideout rates varied with the species and with bulk water concentration. Field evaluations of crevice impurity inventory models based on the hideout rate data indicated that further model refinements were necessary, e.g., more frequent quantification of the relation of hideout rates and bulk water concentration. An alternate crevice inventory model based on a real-time mass balance approach also began to be pursued. Modeling results currently are being used at several PWRs to establish a chloride injection rate consistent with development of a near neutral crevice solution to minimize IGA/SCC. Hideout return data are being used to independently establish predictions of crevice chemistry and to substantiate the hideout rate and mass balance model predictions

  13. The underclad cracking in PWR reactor vessels

    International Nuclear Information System (INIS)

    The article describes the kind of cracking which can occur under the stainless steel cladding during the manufacturing process of PWR vessels: - cold cracking recently found in France on vessel nozzles-reheat cracking discovered some ten years ago in particular in Germany and in USA. Methods of examination for underclad cracking are put forward, together with results obtained on vessel nozzles of units currently being built in Belgium. Some nozzles are affected by the phenomenon of reheat cracking, whilst the hypothesis of cold cracking, which had been proposed because of the similar situation found in France should probably be abandoned. On the basis of the investigations and studies made, it is established that the cracking involved does not jeopardize the integrity of the vessels during their life time. (author)

  14. The material analysis for PWR primary equipment

    International Nuclear Information System (INIS)

    The primary equipment in pressurized water reactor includes reactor pressure vessel, reactor coolant piping, steam generator, pressurizer, and reactor coolant pump casing, etc., which form the pressure boundary of the primary loop. These primary equipment are all pressure vessels of QA Class 1, Safety-related Class 1, and Aseismatic Category 1. Under high temperature, high pressure and neutron irradiation, the requirements for the base material and welding properties of these pressure vessels are very high, so as to ensure the long-term stable operation of nuclear power plant. The base material and welding properties of these pressure vessels are analyzed and discussed according to ASME B and P Code, which can be as a reference for base material selection of PWR pressure vessels. (authors)

  15. The Regulatory System for the First NPP in Indonesia

    International Nuclear Information System (INIS)

    The first Nuclear Power Energy Resources (NPP) will operate in 2016 in order to fulfill the needs of national electrical energy especially in Java, Madura and Bali (JAMALI), It was stated in the blue print in National Energy Management 2005 - 2025. The regulatory aspects of the use of nuclear energy in accordance with the role and functions of BAPETEN (Nuclear Energy Regulatory Agency) are to prepare and develop rules and regulations, licensing system, inspection system, as well as human resources. The road map in the regulation of the first NPP has been in place since 2004. According to the challenges in the Strategic Plan of BAPETEN, BAPETEN carries out the regulatory program for the first NPP, for the ageing management of research reactors, as well as for nuclear security. In this paper, I will present the preparation of the regulatory aspects of NPP according to the role and functions of BAPETEN, i.e. to prepare and develop rules and regulations, licensing system, inspection system, as well as human resources. Rules and regulations on NPP has been enacted and some are still being prepared and developed. Government Regulation No. 43 of 2006 regulates the licensing of nuclear installations, including site license, construction license, operation license, commissioning license and decommissioning license. Government Regulation that is being drafted is on Nuclear Safety and Security. Presidential Regulation on Nuclear Liability is expected to be enacted by the end of this year. The licensing system prepared in facing the operation of the first NPP in Indonesia should begin with the preparation of the process and procedures for obtaining NPP licenses. The review and assessment procedures in every stage of NPP licenses (site, construction, commissioning, operation and decommissioning) should be prepared in line with the existing regulations. The inspection and enforcement procedures in the construction and operation of NPP are prepared based on the applicable rules

  16. Characterization of Factors affecting IASCC of PWR Core Internals

    Energy Technology Data Exchange (ETDEWEB)

    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. The PWR cores management; La gestion des coeurs REP

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  18. Hydraulic benchmark data for PWR mixing vane grid

    International Nuclear Information System (INIS)

    The purpose of the present study is to present new hydraulic benchmark data obtained for PWR rod bundles for the purpose of benchmarking Computational Fluid Dynamics (CFD) models of the rod bundle. The flow field in a PWR fuel assembly downstream of structural grids which have mixing vane grids attached is very complex due to the geometry of the subchannel and the high axial component of the velocity field relative to the secondary flows which are used to enhance the heat transfer performance of the rod bundle. Westinghouse has a CFD methodology to model PWR rod bundles that was developed with prior benchmark test data. As improvements in testing techniques have become available, further PWR rod bundle testing is being performed to obtain advanced data which has high spatial and temporal resolution. This paper presents the advanced testing and benchmark data that has been obtained by Westinghouse through collaboration with Texas A&M University. (author)

  19. Mochovce NPP unit 1 and 2 completion project

    International Nuclear Information System (INIS)

    History of Mochovce NPP is quite long and reflects political changes that happened in Europe in the end of 80'ties. The plant site was chosen in south-west of Slovakia in the frame of Nuclear Industry Development Plan adopted by former Czechoslovak government in 70'ties. In that time was decided to build in Mochovce four VVER 440/213 units together with other NPP's (Dukovany, Temelin, Kecerovce)

  20. Morbidity with temporary disability in Kozloduy NPP workers

    International Nuclear Information System (INIS)

    Changes with time in indicators of disease incidence with temporary disability in Kozloduy NPP personnel have been studied for the period 1974-1991. The data were compared with those for 'Sofia-Iztok' TPP. The causes contributing to formation of the indicators of frequency, severity, and average duration were examined. No temporary disability because of radiation exposure has been recorded. As a whole, less temporary disability has been found at NPP than at TPP. (author)

  1. Students education and training for Slovak NPP

    International Nuclear Information System (INIS)

    elaboration of diploma thesis. In addition to regular academic education we perform post-gradual courses: 'Safety aspects of NPP operation'. The main goal is to increase safety culture of NPP operation and target groups are operation staff of NPP, NRA officers, nuclear safety specialists - all graduated from technical universities with at least two years practice in nuclear industry. On international level we organised the 4 weeks 'IAEA Regional Training Course on Safety, Management and Utilization of Research Reactors' which was held in Bratislava (Slovakia) and Vienna (Austria) during March 05-30th 2001. IAEA in co-operation with the Department of Nuclear Physics and Technology of the Slovak University of Technology and the Atominstitut of Austrian Universities Vienna prepared and realized this training course with the aim to train junior staff from research reactors in various aspects of safety, management and utilization of research reactors. All participants had to have at least 4 years experiences in operation, management, utilization or regulation of research reactors. Lectures covered the topics in nuclear design and operation, neutron physics, reactor physics, health physics, dosimetry, reactor instrumentation, fuel management decontamination procedures, preparation of experiments at research reactors and others. Beside theoretical part of the course, the practical exercises at TRIGA II reactor in Vienna constituted an important part of training. The course was held in English for participants from 6 countries (Slovakia, Russia, Romania, Hungary, Ukraine, Turkey) and thank support of IAEA was fully provided with textbooks and laboratory guides. This year we take part via students and 2 professors the second run of the Eugene Wiegner course establishing in frame of ENEN project. According to international experiences obtained during the last 3 years, we created The Slovak Nuclear Education Network (SNEN) which is supervised at our Department. Coordination of nuclear

  2. Hot Operation of FTL for PWR Fuels Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sung Ho; Joung, Chang Yong; Lee, Jong Min; Park, Su Ki; Sim, Bong Sik; Ahn, Guk Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Fuel Test Loop (FTL) in HANARO is the test facility which can conduct a fuel irradiation test with commercial NPPs' operating conditions such as their pressure, temperature, flow and water chemistry. The FTL is used for the irradiation test of PWR type or CNNDU type fuels. In this paper, the hot operation of FTL for irradiation test of PWR fuels is introduced. The experimental results show the excellence of operation performance

  3. Human Resources Training Requirement on NPP Operation and Maintenance

    International Nuclear Information System (INIS)

    This paper discussed the human resources requirement on Nuclear Power Plant (NPP) operation and maintenance (O&M) phase related with the training required for O&M personnel. In addition, this paper also briefly discussed the availability of training facilities domestically include with some suggestion to develop the training facilities intended for the near future time in Indonesia. This paper was developed under the assumptions that Indonesia will build twin unit of NPP with capacity 1000 MWe for each using the turnkey contract method. The total of NPP O&M personnel were predicted about 692 peoples which consists of 42 personnel located in the head quarter and the rest 650 people work at NPP site. Up until now, Indonesia had the experience on operating and maintaining the nonnuclear power plant and several research reactors namely Kartini Reactor Yogyakarta, Triga Mark II Reactor Bandung, and GA Siwabessy Reactor Serpong. Beside that, experience on operating and maintaining the NPP in other countries would act as one of the reference to Indonesia in formulating an appropriate strategy to develop NPP human resources particularly in O&M phases. Education and training development program could be done trough the cooperation with vendor candidates. (author)

  4. EPC projects for EPR Flamanville 3 NPP

    Energy Technology Data Exchange (ETDEWEB)

    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

  5. Lessons learnt from Ignalina NPP decommissioning project

    International Nuclear Information System (INIS)

    The Ignalina Nuclear Power Plant (INPP) is located in Lithuania, 130 km north of Vilnius, and consists of two 1500 MWe RBMK type units, commissioned respectively in December 1983 and August 1987. On the 1. of May 2004, the Republic of Lithuania became a member of the European Union. With the protocol on the Ignalina Nuclear Power in Lithuania which is annexed to the Accession Treaty, the Contracting Parties have agreed: - On Lithuanian side, to commit closure of unit 1 of INPP before 2005 and of Unit 2 by 31 December 2009; - On European Union side, to provide adequate additional Community assistance to the efforts of Lithuania to decommission INPP. The paper is divided in two parts. The first part describes how, starting from this agreement, the project was launched and organized, what is its present status and which activities are planned to reach the final ambitious objective of a green field. To give a global picture, the content of the different projects that were defined and the licensing process will also be presented. In the second part, the paper will focus on the lessons learnt. It will explain the difficulties encountered to define the decommissioning strategy, considering both immediate or differed dismantling options and why the first option was finally selected. The paper will mention other challenges and problems that the different actors of the project faced and how they were managed and solved. The paper will be written by representatives of the Ignalina NPP and of the Project Management Unit. (author)

  6. Regulation and license for NPP in Indonesia

    International Nuclear Information System (INIS)

    Any activity relating to nuclear energy shall maintain safety, security, safeguards, health worker and public as well as environmental protection, according to article 16(1) act No. 10 year 1997 on Nuclear Energy. The act No. 10 year 1997 on Nuclear Energy stipulated independent government agency as nuclear energy regulatory called BAPETEN (Nuclear Energy Regulatory Agency). BAPETEN has task to make regulation, licensing process and inspection. Bapeten shall have a licensing and inspection system in fulfilling his function to issued licensing according to Nuclear Energy Act. In Article 17 stipulated that any nuclear energy utilization or construction and operation of nuclear reactor and other nuclear installation as well as decommissioning of nuclear reactor shall have license. Requirements and process to issue license for nuclear reactor stipulated further in Government Regulation no. 43 year 2006. The purpose of the Government Regulation No. 43 is to stipulate licensing of nuclear reactor in order to assure the health of worker and public, environment protection and security of nuclear material and nuclear facility. Requirement and guidance in licensing NPP mandated by Government Regulation or Presidential Decree stipulated in Bapeten Chairman Regulation

  7. Management of occupational exposure at Cernavoda NPP

    International Nuclear Information System (INIS)

    Ionising radiation presents a particular risk associated with nuclear power plant operation. An effective and efficient radiation protection program should: - prevent the detriment of health due to deterministic effects; - keep all the exposures as low as reasonably achievable in order to limit the detriment of health due to stochastic effects; - provide safety and health conditions as good as other safe industries. Radiation protection of occupationally exposed workers is part of Health and Safety of Work Program. Effective dose limits, as recommended by ICRP and required by CNCAN regulations are reasonably low in order to avoid deterministic effects and to limit the probability of stochastic effects to an acceptable level. The health status of Cernavoda NPP employees is appropriately surveyed. There were not recorded cases of occupational diseases and/or other indicators of relevant biological effects in order to establish the specific response of the human body to the occupational illness risk factors. Starting in 2002 the cytogenesis blood analysis for occupationally exposed individuals was performed at the beginning of their employment and then periodically for those working for more than five years in the plant. Up to 1900 employees have been investigated with no indication of genetic modifications. (author)

  8. NPP lifetime philosophy: the transatlantic difference

    International Nuclear Information System (INIS)

    Fundamental institutional and cultural differences in the transatlantic nuclear power industries, and in particular those between the Nordic countries and the United States, have driven divergent plant life management strategies -strategies resulting in distinctly different plant performance. Recognition of the linkage between three key components of overall Nuclear Power Plant (NPP) performance - yearly O and M costs, safety, and effective plant lifetime -is based on different institutional perspectives. In the Nordic countries, explicit recognition of this linkage has been historically translated into an integrated approach to plant performance. American NPPs, however, have been forced to focus primarily on near term O and M performance and regulatory mandated investment. While Nordic NPPs view capital investment in plant lifetime management and modernization as necessary to avoid declining plant performance and the cost of replacement power, American NPPs exhibit reluctance for such investments due to the difficulty of justifying the associated short-term costs. The diverging histories of two NPPs of the same vintage and design, one in Sweden and one in the United States, exemplify the potential ramifications of these approaches. The Swedish plant continues to operate with excellent performance indicators, while undertaking a comprehensive and long-term modernization program. The American facility is likely to be decommissioned due to unsustainable economic performance. (author)

  9. Training needs analysis at Cernavoda NPP - Romania

    International Nuclear Information System (INIS)

    The paper is mainly on training needs analysis applied at Cernavoda NPP and will outline the Cernavoda (CNPP) approach in establishing what kind of training is necessary for CNPP employees, from the moment they are hired until they become qualified. The training methodology adapted at CNPP is one which adheres to the principles of the Systematic Approach to Training (SAT). SAT adoption at CNPP provides a broad integrated approach emphasizing not only technical knowledge and skills but also human factors knowledge, skills and attitudes. The analysis of SAT at CNPP consists of a 'table-top' analysis of training and qualification requirements. This paper will illustrate how a 'table-top' analysis by Subject Matter Experts is organized, carried out and recorded, and which categories of staff have begun this table-top analysis at CNPP. It will also give an example how this analysis is done: how the tasks of each job are rated, which are the used criteria, and how to proceed with the incorporation of the tasks into initial and continuing training, as appropriate. (author)

  10. SAT for NPP personnel training in Spain

    International Nuclear Information System (INIS)

    The SAT process objectives as applied to Spanish NPPs are: Perform JTA for selected job positions at the NPPs; develop the associated training plans; develop training support media; prepare training instructors as teachers and as task analysts; develop a SAT database. A lesson-learned from the task assessments conducted at Spanish NPPs is that the final task list should be obtained with full participation of workers so that they feel the final training plan has been developed taking into account their own opinion and experience. The breakdown of the tasks into its elements and the concordant job performance measure was presented. The process of determining knowledge and skills associated with the task elements and the structure and use of the taxonomy codes (for component, system, and academic skills and knowledge) was shown and explained. Based on the Spanish experience, the average time devoted to analyze a complete task is 5.33 person-hours per task. This data has permitted training and plant management to allocate the NPP human resources to support the analysis phase of SAT (which can be very time-consuming)

  11. Thermal performance monitoring in Dukovany NPP

    Energy Technology Data Exchange (ETDEWEB)

    Pliska, J.; Machat, Z.; Horky, V.; Sury, P.; Havlat, L. [I and C Energo plc., Prazska 684, 674 01 Trebic (Czech Republic)

    2010-07-01

    The paper presents a practical application of the thermal performance monitoring on Dukovany NPP related to the performance of the balance of plant. The system described is focused on the health of the process and on the health major equipment involved in the process. The system provides information on actual and expected or desirable (should-be) values of thermal performance indicators for performance engineers, maintenance planners and control room operators. The system is designed to monitor steady states. The system contains two main functions: Data validation and Process simulation. Data validation provides a solid base for detection, identification and elimination of process data corrupted by gross errors and for monitoring and evaluation of measurement degradation. Process simulation is performed with steady state mathematical model of power plant components configured into secondary side. Major purpose of Process simulation is calculation of performance data for an expected state of the plant equipment and the actual mode of operation. Control room operators (in near-real time) and performance engineers then can evaluate and diagnose the health of the equipments and the adjustment of the process parameters by comparing an actual and expected state of the performance data. The system also allows 'what-if' analysis and efficiency optimization based on varying user selected parameters of the plant components, process parameters and flow-sheet structure changes. (authors)

  12. Economic aspects of Dukovany NPP fuel cycle

    International Nuclear Information System (INIS)

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

  13. Intranet portal at the Krsko NPP, Slovenia

    International Nuclear Information System (INIS)

    The intranet portal (named IntraNEK) at Krsko NPP serves as a single entry point to access the internet and various plant applications and links. The front page consists of the standard internet search bar and links to various applications that can either reside within the technological computer network (TRM) or within the plant business computer network. Access to the TRM applications is read only. Some applications on the business computer network are open to all personnel who log on to the network while some applications are restricted and secured, and require additional login entries. A selected link will open in a new window. Documents will open with the appropriate software tool depending on the document file format. Some categories of documents are available in image form only (e.g. procedures, drawings etc.), while some are available in fully searchable PDF format (e.g. technical specifications, updated safety analysis reports (USARs) etc.). Plant departments (organizational units) have their own pages accessible from the front page. Their pages contain links to their own information resources or links to other resources and applications, tailored to the department needs. During recent years a number of web based applications have been developed that are connected also with a common Oracle database. Some are designed to serve for data entry and browsing while others serve for browsing only

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  15. Seawater desalination using reusable type small PWR

    Energy Technology Data Exchange (ETDEWEB)

    Uchiyama, Y. [Institute of Engineering Mechanics and Systems, University of Tsukuba, Tsukuba, Ibaraki (Japan); Minato, A. [Planning Division, Central Research Institute of the Electric Power Industry, Komae-shi, Tokyo (Japan); Shimamura, K. [Nuclear Systems Engineering Department, Nuclear Energy Systems Engineering Center, Mitsubishi Heavy Industries, Ltd., Kanagawa (Japan)]. E-mail: shimamura@atom.hq.mhi.co.jp

    2003-07-01

    Demand for seawater desalination is increasing, especially in regions such as the Middle East and North Africa, where populations are growing at a high annual rate. If such demand is met by fossil fuel energy, the influence on the environment, such as global warming, cannot be disregarded. Since these regions are behind in their preparedness of social capital infrastructure, such as power transfer grids, small reactors are considered to be more suitable for introduction than the large reactors found commonly in developed countries. Therefore, a small reusable PWR with mid-range pressure and temperature services, which does not require on-site refuelling, was devised for seawater desalination. In a small reusable PWR, spent fuel is taken out together with the reactor vessel and refuelled on the exterior fuel exchange base prepared independently. Thus, the safeguards against nuclear proliferation increase at a plant site because the lid of the reactor vessel is never opened at the site, in principle. The reactor vessel will be transported from the plant site to a fuel exchange base under stipulated conditions within a transportation cask after a long (about six years) operation. Since fuel handling facilities at the site become unnecessary through centralisation at a fuel exchange base, initial plant construction costs are reduced. In addition, the reactor vessel is reused until its service life has expired. This examination was based on the marine reactor of the experimental nuclear ship, Mutsu, after it had been applied for land use: at a lowered, midrange pressure and temperature service, in theory. It is possible to produce fresh water through reverse osmosis (RO) membrane pressure-rising seawater by a steam turbine driven pump. Using the method of driving a desalination unit high-pressure pump directly by low-pressure steam generated from the heating reactor, fresh water can be produced efficiently. Furthermore, operating at reduced pressure makes it possible

  16. Validation and verification of a deterministic model for corrosion and activity incorporation using operational data from Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Betova, I. [Technical Univ. of Sofia, Sofia (Bulgaria); Bojinov, M. [Univ. of Chemical Technology and Metallurgy, Sofia (Bulgaria); Minkova, K. [Kozloduy Nuclear Power Plant, Kozloduy (Bulgaria)

    2010-07-01

    An integrated deterministic model of corrosion and activity incorporation in the construction materials of the primary circuit of light water reactors based on fundamental physico-chemical processes has been recently proposed. The calculational procedure of the model enables to obtain reliable estimates of the kinetic and transport parameters of growth and restructuring of inner and outer oxide layers on austenitic steels (AISI 304, 0X18H10T, AISI 316, A800) and nickel alloys (A600 ∦ A690) via a quantitative comparison of the model equations with electrochemical data on the conduction mechanism and ex-situ analytical information on the thickness and in-depth chemical composition of the oxides on such materials stemming from both laboratory and in-pile BWR, PWR and WWER reactor experience. As a result, a large database of kinetic and transport parameters makes it possible to predict the kinetics of growth and restructuring of the oxides, as well as corrosion release from the construction materials in the primary circuit, by using data on the water chemistry and radioactive corrosion products in the coolant and the piping surfaces. Predictions on the incorporation of radioactive corrosion products during oxide growth and restructuring are also made. In the present communication, the validation and verification of the model using data for the primary circuit water chemistry and radioactive corrosion products from both reactors of the Kozloduy NPP are discussed. The calculations are in good agreement with the available experimental data and allow for reliable long-term predictions of the corrosion processes and radioactivity accumulation in the primary circuit of both reactors of Kozloduy NPP to be obtained. (author)

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

    DEFF Research Database (Denmark)

    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...

  18. Achievements and challenges of Paks NPP

    International Nuclear Information System (INIS)

    As the six year long safety upgrading program at Paks NPP is approaching its final stage this year, it is a good opportunity to draw the conclusion: what have been done and how have measures influenced the safety of the plant. In its first part the paper gives an overview of the program's main issues, assesses the results from the point of view of safety, reliability and cost effectiveness as well. In the second part a survey of future tasks follows: (1) Hungary is joining to the EU. The accession process so far has not revealed any major problems concerning nuclear safety which could be seen as obstacles toward the membership. However the plant should be ready to meet the increasing level of safety regulations. Further safety upgrading measures are planned, mostly in the field of severe accident management. (2) The electricity market liberalisation in Hungary will start in 2003 and being a EU member state, the full market opening will happen within a few years. The plant has to take into account the specificity of market functioning. The most important thing is to preserve the present cost advantage of nuclear electricity generation within the market conditions. The paper presents measures performed and planned to keep the unit generation cost competitive. (3) The first unit at Paks will mark its 20't'h anniversary this year. Lifetime management issues are at the centre of the engineering activities. The work already started to prepare the lifetime extension for 20 years. The program for the license renewal, which was elaborated jointly with the nuclear regulatory body will be described.(author)

  19. Presentation of Gentilly 2 NPP containment

    International Nuclear Information System (INIS)

    The paper (abstract only) describes the plant significant features and the concrete containment life management including ageing related problems and remedies applied. Gentilly 2 NPP is a PHWR CANDU 6 nuclear power plant with a net 600 MW capacity, located in Quebec (Canada). During 15 years operations, the plant was shutdown every year for maintenance and repairs including the periodic reactor building leak rate test. The containment is a 35 Mpa concrete cylinder with a dome roof, resting on a concrete base slab. It is reinforced and pre-stressed structure with an inside diameter of 41.4 meters and a wall thickness 1070 mm for the cylinder wall and 610 mm for the dome. BBR was selected as pre-stressing system, with independent cables in the containment base slab, cylinder and dome. Strain gauges were embedded in the concrete with readings in 2 directions at each location, and can provide an insight into the concrete ageing process at each subsequent high pressure leak rate test. The variation of the leak rate, a sign of ageing, is discussed in this paper, as well as the remedies applied to reduced it. The non-conformances during construction and which had an impact on the containment ageing process are also discussed. Finally the ageing management program is described, including inspection and testing. Under testing are presented the results of the periodic high pressure leak rate tests performed and the low pressure test performed at full power, of the lift-off and of the test beams tests, as part of monitoring of the ageing of the concrete and pre-stressing cables respectively. Where remedial actions were applied, the method used and results obtained are described

  20. Spent fuel management of Jose Cabrera NPP

    Energy Technology Data Exchange (ETDEWEB)

    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)

  1. The Study of Nuclear Fuel Cycle Options Based On PWR and CANDU Reactors

    International Nuclear Information System (INIS)

    The study of nuclear fuel cycle options based on PWR and CANDU type reactors have been carried out. There are 5 cycle options based on PWR and CANDU reactors, i.e.: PWR-OT, PWR-OT, PWR-MOX, CANDU-OT, DUPIC, and PWR-CANDU-OT options. While parameters which assessed in this study are fuel requirement, generating waste and plutonium from each cycle options. From the study found that the amount of fuel in the DUPIC option needs relatively small compared the other options. From the view of total radioactive waste generated from the cycles, PWR-MOX generate the smallest amount of waste, but produce twice of high level waste than DUPIC option. For total plutonium generated from the cycle, PWR-MOX option generates smallest quantity, but for fissile plutonium, DUPIC options produce the smallest one. It means that the DUPIC option has some benefits in plutonium consumption aspects. (author)

  2. The terrestrial NPP simulations in China since Last Glacial Maximum

    Institute of Scientific and Technical Information of China (English)

    HE Yong; DAN Li; DONG Wenjie; JI Jinjun; QIN Dahe

    2005-01-01

    Based on Atmosphere-Vegetation Interaction Model (AVIM), the magnitude and spatial distribution of terrestrial net primary productivity (NPP) in China is simulated during three different geological eras, Last Glacial Maximum (LGM), Mid-Holocene (MH) and the present. The simulation shows that the glacial-interglacial variation of East Asian summer monsoon in China is the key factor affecting the NPP change. During the three eras, mean NPPs are 208 g/m2·a, 409 g/m2·a, and 355 g/m2·a. The total NPPs are 2.05 Pg/a, 3.89 Pg/a and 3.33 Pg/a, respectively. The terrestrial NPP in China during warm-humid climate is larger than that during cold-arid eras, and the correlation analysis between NPP and climate factors suggests that temperature is the primary factor affecting the terrestrial NPP during 21 kaBP (LGM), and for 6 kaBP (MH) and the present the primary factor is precipitation.

  3. New appraisement of siting for a NPP on Mures river

    International Nuclear Information System (INIS)

    The studies for a second NPP siting on inner Romanian rivers began in a careful manner since 1982 as a first part of the Nuclear Power Plant Romanian Program adopted by political and governmental authorities at the time. The experience gained from Cernavoda NPP siting, the first mission of new multi-branch of specialists team was to choose new NPP sites adapting the CANDU type NPP Cernavoda project to the new parameters of close water cooling circuit and of hard less or no rock foundation strata. The new sites conditions mean a lot of changes of CANDU license and a decrease the output power supplied to the national electric grid. The studies on the Mures river as alternative site of Olt river in Transylvania region began in 1986 and were stopped after 1990. This paper tries to reconsider shortly the old analysis focused on geological and geotechnical aspects and other local sites characteristics according to the last IAEA Safety Standards taking into account also the last types of NPP generations and the number of units. (author)

  4. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  5. Maintenance technologies for SCC of PWR

    International Nuclear Information System (INIS)

    The recent technologies of test, relaxation of deterioration, repairing and change of materials are explained for safe and stable operation of pressurized water reactor (PWR). Stress corrosion cracking (SCC) is originated by three factors such as materials, stress and environment. The eddy current test (ECT) method for the stream generator pipe and the ultrasonic test method for welding part of pipe were developed as the test technologies. Primary water stress corrosion cracking (PWSCC) of Inconel 600 in the welding part is explained. The shot peening of instrument in the gas, the water jet peening of it in water, and laser irradiation on the surface are illustrated as some examples of improvement technology of stress. The cladding of Inconel 690 on Inconel 600 is carried out under the condition of environmental cut. Total or some parts of the upper part of reactor, stream generator and structure in the reactor are changed by the improvement technologies. Changing Inconel 600 joint in the exit pipe of reactor with Inconel 690 is illustrated. (S.Y.)

  6. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    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

  7. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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 10−6 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. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code

    International Nuclear Information System (INIS)

    When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

  13. Method and result of experiment for support of technical solutions in the field of perfection of a nuclear fuel cycle for future PWR reactors

    International Nuclear Information System (INIS)

    This paper represents conceptual approaches of statement and carrying out of experiments to validate functional safety of PWR reactors of the future, at acceptance of technical solutions on use of fuel rods with the increased length of a fuel column in fuel assemblies. The paper represents main principles and criteria, which we use for quality check of technical solutions and developments in the field of perfection of a nuclear fuel cycle of PWR reactors of the future, first of all, from the point of view of a substantiation of safety of the future operation at change of fuel rod design. We explore the safety issues of operation of PWR reactors with fuel assemblies, including fuel rods with various length of a fuel column. The paper discusses the ways of solving of the important problems of carrying out of critical facility experiments for verification of new technical solutions in the field of PWR nuclear fuel cycle improvement on the base of international standards ISO 2000:9000 and functional safety recommendations on functional safety of IEC (International Electromechanical Commission). The package of new Federal Laws of the Russian Federation in the field of safety and licensing of activity of dangerous manufactures defines a major principle for requirements to the supplier of nuclear techniques and NPP as a whole. This principle is - for any moment of operation of NPP quantity indicators of risk should not exceed comprehensible social size of the established indicators of safety. On the other hand the second principle should be applied from operation of the equipment, systems or NPP as a whole to extraction of the greatest benefit: As much as possible long operation and full commercial use of resource and service properties of the equipment, systems and NPP as a whole. Realization of this principle assumes development and introduction of new technical solutions for a validation of guarantees of safety of the future operation of NPP or it separate components

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

    Science.gov (United States)

    Volmert, Ben; Pantelias, Manuel; Mutnuru, R. K.; Neukaeter, Erwin; Bitterli, Beat

    2016-02-01

    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.

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

    Directory of Open Access Journals (Sweden)

    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.

  16. Use of control room simulators for training of NPP personnel

    International Nuclear Information System (INIS)

    The objective of this report is to provide NPP managers, training center managers and personnel involved with control room simulator training with practical information they can use to improve the performance of their personnel. While the emphasis in this report is on simulator training of control room personnel using full scope simulators, information is also provided on how organizations have effectively used control room simulators for training of other NPP Personnel, Vienna (AT) including simulators other than full-scope simulators. The documents includes: the main body with current practices and recommendations; selected examples from countries; a CD ROM with all examples (different languages). The document will be available on the IAEA web site. The topics describes are: trends in simulators training; designing and developing training involving room simulators; implementation of simulator training; evaluating the effectiveness of simulator training; simulator instructor competence; application of different types of simulators in the training of NPP personnel (other than full scope simulators

  17. Main principles of the Chernobyl' NPP zone development

    International Nuclear Information System (INIS)

    It is suggested to divide the Chernobyl' NPP zone into two parts, which are the alienation and evacuation (buffer) zones. The alienation zone includes the areas with greatest contamination around the Chernobyl' NPP. The population residence in this zone is forbidden. The watching method of working with short-time personnel residence is suggested to be used in this zone. The buffer zone is the territory out of the alienation zone boundaries including all settlements, from which the population is evacuated. Constant residence is permitted in the buffer zone for persons 50 and more years old with introduction of restrictions for diet and residence organization. The production activity in this zone includes operation of three units of the Chernobyl' NPP, works with the Ukrytie object and researches. Operations connected with radioactive waste processing and redisposal from places of storage is not recommended to be done. It is suggested to develop methods for local radioactive waste processing

  18. Main public health risk factors nearby NPP construction site

    International Nuclear Information System (INIS)

    Full text:This paper discusses methods, which were developed based on original results and international recommendations, for processing and analysis of data on environmental monitoring to assess risk of adverse ecological consequences. The methods considered are also applicable for the data relevant to the territories nearby NPP construction site. Factor of both radiational and non-radiational origin, which may cause adverse health effects, were analyzed. An algorithm for assessment of environmental contamination was developed and coded in Delphi. The algorithm allows calculating integral contamination coefficient, getting information on safety state of a facility, and studying human body interaction with the environment under various conditions at work and home. Existing chemical and radiational parameters of the environment in the vicinity of the NPP construction site were considered. A potential environmental contamination form radioactive gaseous and aerosol emission during normal operation of 2 VVER-1200 units was estimated. Calculations showed that maximal soil surface contamination will not exceed 0.144 Bq/m2 (3.89·10-6 Ci/km2) during the first year of NPP operation. By the end of 60 years of NPP operation, maximal soil contamination will have increased by 1.67 Bq/m2 (4.5·10-5 Ci/km2). Fraction of soil contamination with radionuclides from NPP emission will be insignificant and will have reached 9.0·10-2 % by the end of 60 years of NPP operation. Individual annual exposure of population due to gaseous and aerosol emissions is 4.5·10-4 mSv (0.045% of the annual dose limit as recommended by radiation safety norms) in the maximal contamination spot at normal operation of 2 VVER-1200 units. Experimental measurements of pollutants' content in the atmospheric air on the territory being researched were done at Republic Center for Radiation Control and Environmental Monitoring. Experimentally obtained data on radioactive contamination of atmospheric air nearby

  19. Comprehensive radiation survey and public health assessment in the vicinity of NPP

    Directory of Open Access Journals (Sweden)

    Samoilov A.S.

    2015-12-01

    health indicators of the public living around the NPP, general state of the adult and children public is healthy. There is no impact of the NPP on negative trends in some changing indicators of the adult and children public health.

  20. Romanian contribution to the design of Cernavoda NPP

    International Nuclear Information System (INIS)

    The Cernavoda NPP is the first nuclear power unit built in Romania. It resulted as implementation of the national nuclear program provisions worked out in the period 1960-1978. The Center of Technology and Engineering for Nuclear Projects operated in the Cernavoda NPP construction in its capacity of general and specialty designer. It conducted studies and design works to select optimal nuclear reactor system, nuclear fuel cycle and associated nuclear materials (D2O, Zr), aiming particularly to the introduction of CANDU type power reactors in the Romanian power system. In the period 1977-1990 basic technical and economical documentation were worked out concerning the site of construction, technical annexes to the commercial contracts, procurement order documents, technical assistance on site of construction, design for equipment fabrication, etc. In 1990-2002 significant changes in conducting the project have taken place. In the period August 1991- June 1996 the project management was taken over by the ACC consortium formed of AECL (Canada) and ANSALDO (Italy), while CITON was acting as principal contractor of ACC for design services and technological engineering. The paper presents the following main achievements in the design work: site selection and NPP general layout; nuclear safety and environmental protection; structure and infrastructure works; anti-seismic design; design of process systems; thermomechanical equipment for the NPP nuclear sector; design for electrical systems and automation. In conclusion, the contribution to the design and construction of the Romanian specialists for realization of Cernavoda NPP Unit 1 was decisive, while the cooperation with Canadian and Italian teams was extremely fruitful. These constitute a valuable potential to be used in building the Cernavoda NPP Unit 2 and Units 3-5

  1. Modern practice of cost estimates for the NPP units decommissioning

    International Nuclear Information System (INIS)

    The results of analysis of current practices of cost estimates for decommissioning of nuclear power units with different reactor types present is reviewed. Cost estimates intervals are shown for decommissioning of units with PWR,BWR and AP1000 reactors and the main factors influencing the cost amount are analyzed

  2. NPOESS Preparatory Project (NPP) Environmental Products

    Science.gov (United States)

    Grant, K. D.; Smith, D. C.

    2011-12-01

    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). The Joint Polar Satellite System will contribute the afternoon orbit component and ground processing system of the restructured National Polar-orbiting Operational Environmental Satellite System (NPOESS). As such, the Joint Polar Satellite System replaces the current Polar-orbiting Operational Environmental Satellites (POES) managed by NOAA and the ground processing component of both POES and the Defense Meteorological Satellite Program (DMSP) replacement, known as the Defense Weather Satellite System (DWSS), managed by the Department of Defense (DoD). The Joint Polar Satellite System satellite will carry a suite of sensors designed to collect meteorological, oceanographic, climatological, and solar-geophysical observations of the earth, atmosphere, and space. The ground processing system for the Joint Polar Satellite System is known as the Common Ground System (JPSS CGS), and consists of a Command, Control, and Communications Segment (C3S) and the Interface Data Processing Segment (IDPS). Both are developed by Raytheon Intelligence and Information Systems (IIS). The Interface Data Processing Segment will process Joint Polar Satellite System and Defense Weather Satellite System satellite data to provide environmental data products (aka, Environmental Data Records or EDRs) to the National Oceanic and Atmospheric Administration and Department of Defense processing centers operated by the United States government. The Interface Data Processing Segment will process Environmental Data Records beginning with the NPOESS Preparatory Project (NPP) and continue through the lifetime of the Joint Polar Satellite System and Defense Weather Satellite System programs. Under the National Polar-orbiting Operational Environmental

  3. Signal processing methods for PWR reactor noise diagnostic system

    International Nuclear Information System (INIS)

    A framework for a PWR reactor noise diagnostic system using various signal processing methods has been investigated. Supposing to treat not only reactor noise data in a stationary linear system but also those in a nonstationary or nonlinear system, the study covers a third-order-correlation of bispectrum, cepstrum analysis, Group Method of Data Handling (GMDH), chaotic quantity, neural network, and wavelet, in addition to Multivariate AutoRegressive analysis and Signal Transmission Path Diagram analysis (MAR/STPD). This paper describes consideration about the methods from viewpoints of theories and applications to PWR reactor noise diagnostic system. The point at the issue in the application system is how to extract many characteristics from the signals whatever states (linear or nonlinear, stationary or nonstationary) may happen in order to get more information and more exact diagnose to support human judgment. From this viewpoint, the paper discusses several signal processing techniques for the PWR diagnostic system. (J.P.N.)

  4. Industry-wide survey of organics in PWR's

    International Nuclear Information System (INIS)

    Interest in organic impurities found in Pressurized Water Reactors (PWR's) has stemmed from several sources. The most serious concern is that organic acids will increase cation conductivity, a parameter that is used to control power plant chemistry. This effect can complicate secondary water monitoring and control. Organics may foul or exhaust makeup demineralizers and condensate polishers, and thus result in increased operating costs or the in leakage of potentially corrosive agents into the steam generators. Some organics, however, such as mopholine and cyclohexylamine may reduce corrosion through oxygen scavenging or surface filming reactions, and may have a positive influence on the pH in areas of local corrosion. At the time this survey began, little information was available on the types or levels of organic impurities that are typically found in PWR's. this survey is intended to provide baseline data for future corrosion testing and to provide fundamental information that will be helpful in refining PWR chemistry guidelines and operating practices

  5. NPP capital equipment life control facilities and tools

    International Nuclear Information System (INIS)

    Algorithms and simulation package for calculation of eigen modes of coolant pressure oscillations in equipment of primary coolant circuit of WWER-1000 NPP, tuned-circuit Q-factor of an oscillating contour of the coolant, fluctuation frequencies pass band and attenuation factors of were developed. Calculation results for cases of presence and absence of gaseous state radiolysis products in the coolant were obtained. Results analysis makes it possible to disclose operating conditions which result in raised vibrations, and to develop control response in NPP industrial control system for their prevention

  6. Management of Damaged SNF Handling Operations At Paks NPP

    International Nuclear Information System (INIS)

    The issue of handling leaky fuel is one of the crucial issues of nuclear energy. It is directly connected with transportation of leaky spent nuclear fuel (SNF) that is mostly stored in the station's cooling pools. At present damaged spent nuclear fuel (SNF) of thirty VVER-440 spent fuel assemblies (SFA) is loaded into the ventilated canisters of types 28 and 29 and temporary stored in the cooling pool of Paks NPP. The report presents the milestones of preparation and safety justification of the technology for preparing canisters with damaged SNF of Paks NPP for transport to FSUE “PA “Mayak” (Russia) for reprocessing. (author)

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

    Directory of Open Access Journals (Sweden)

    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

  8. Load-following operation of PWR plants

    International Nuclear Information System (INIS)

    The load-following operation of nuclear power plants will become inevitable due to the increased nuclear share in the total electricity generation. As a groundwork for the load-following capability of the Korean next generation PWRs, the state-of-the-art has been reviewed. The core control principles and methods are the main subject in this review as well as the impact of load-following operations on the fuel performance and on the mechanical integrity of components. To begin with, it was described what the load-following operation is and in what view point the technology should be reviewed. Afterwards the load-following method, performance and problems in domestic 900 MWe class PWRs were discussed, and domestic R and D works were summarized. Foreign technologies were also reviewed. They include Mode G and Mode X of Foratom, D and L bank method of KWU, the method using PSCEA of ABB-CE, and MSHIM of Westinghouse. The load-following related special features of Foratom's N4 plant, KWU's plants, ABB-CE's Systems 80+, and Westinghouse's AP600 were described in each technology review. The review concluded that the capability of N4 plant with Mode X is the best and the methods in System, 80+ and AP600 would require verifications for the continued and usual load-following operation. It was recommended that the load-following operation experiences in domestic PWRs under operation be required to settle down the capability for the future. In addition, a more enhanced technology is required for the Korean next generation PWR regardless what the reference plant concept is. 30 figs., 19 tabs., 75 refs. (Author)

  9. Borssele PWR noise: measurements, analysis and interpretation

    International Nuclear Information System (INIS)

    In the Borssele reactor - a 450 MWe PWR - reactor noise measurements have been performed during four fuel cycles. Measurements were made with a set of ex-core neutron detectors, on one occasion an in-core displacement transducer, and with primary coolant pressure sensors. Digital analysis was applied, where the most powerful tool was the computer programme FAST, which computes auto and cross power spectra for all combinations from a set of many simultaneously recorded signals. Analyses of neutronic signals show a reactivity noise peak at 9.2 Hz, core barrel motion peaks at about 12 and 15 Hz, a damped oscillation at about 2 Hz. Results are given for begin and end of each fuel cycle. The r.m.s. value of the low frequency noise appears to depend linearly on the boron concentration over a wide range. Also some results of primary coolant pressure noise are presented, with coherent peaks below 15 Hz and incoherent peaks above. The second part of the paper describes an alternative way of analyzing and interpreting noise spectra, namely attempts to decompose the neutronic power spectra into physical components, using the information present in the CPSD's of all detector combinations. The components are characterised by their detector position dependency. Effects considered are: uncorrelated noise, global reactivity noise, core motion attenuation noise, and a possible coupling between reactivity and core motion. Results show excellent separation into reactivity and core motion components with possibilities to separate overlapping peaks. Weak peaks become more easily detectable. At low frequencies the decomposition of the spectra is not yet complete, however. (author)

  10. Development of CFD methodology for investigating thermal-hydraulic characteristics in a PWR dome

    International Nuclear Information System (INIS)

    Highlights: • This study develops a detailed CFD model for the dome of Maanshan NPP. • Flow and heat transfer features in the upper plenum and dome are captured. • Leakage flow to the dome cannot be neglected in the nuclear safety analysis. • Higher EDY and RIY are obtained using the calculated temperature on the RPV head. • It is conservative to take the cold-leg temperature to estimate the EDY and RIY. - Abstract: This study aims to develop a detailed computational fluid dynamics (CFD) model to investigate the flow and heat transfer characteristics in the dome of a pressurized water reactor (PWR). The upper plenum is also considered in order to simulate the possible coolant leak to the dome via the gaps of upper support plate. The essential solid components within the solution domain, including the upper core plate, the guide tube assemblies, the support columns, and the rod cluster control, are realistically modeled, instead of the porous-medium approximation. Through the detailed-geometry CFD simulation, the thermal-hydraulic features in the upper plenum, individual guide tube assembly, and the dome can be obtained. And, the temperature distribution on the reactor pressure vessel (RPV) head can be used to estimate the values of total effective degradation years (EDY) and reinspection years (RIY) for monitoring the crack initiation and growth on the head. Present calculated results also reveal that the original values of EDY and RIY using the cold-leg temperature as the head temperature by the Maanshan staff is conservative

  11. Development of technical risk communication guideline in Nuclear Power Plant (NPP)

    International Nuclear Information System (INIS)

    The way Risk Communication in technological field like NPP should be outlined and executed according to basic principles has not been well described. Therefore a Technical Risk Communication (TRC) guideline was developed to provide directions about key issues of communication between the technical personnel and PSA experts. Applying TRC in NPP can have several benefits like enhancing safety culture in NPP. (author)

  12. Multi-scale evaluation of ISIMIP biome models against NDVI and MODIS NPP data

    Science.gov (United States)

    Rafique, Rashad; Zhao, Fang; Zeng, Ning; Asrar, Ghassem; Reyer, Christopher; Ostberg, Sebastian; Francois, Louis; Tian, Hanqin; Chnag, Jinfeng; Nishina, Kazuya

    2016-04-01

    The net primary productivity (NPP) is commonly used for understanding the dynamics of terrestrial ecosystems and their role in carbon cycle. The global NPP, highly variable over space and time, cannot be directly observed, therefore, satellite based observations of Normalized Difference Vegetation Index (NDVI) are used as a proxy to understand and monitor the NPP dynamics. In this study, we used a combination of most recent NDVI and modeled NPP data for the period 1982-2012, to study the role of terrestrial ecosystems in carbon cycle under the prevailing climate conditions. We found that in general there is good agreement between the spatial patterns and global seasonal cycles between observed NDVI and modeled NPP values. Simulated NPP values also generally agree with MODIS NPP spatially, and temporally, MODIS NPP falls within the model spread of NPP values. Despite of the general agreement in the trends of global total NDVI, MODIS NPP and modeled NPP, considerable spatial differences are found, and the ensemble mean of the models often agrees better with the spatial patterns of observed NDVI and MODIS NPP than individual models.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. The traveller: a new look for PWR fresh fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    Bayley, B.; Stilwell, W.E.; Kent, N.A. [Westinghouse Electric Co., Columbia, SC (United States)

    2004-07-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain.

  15. The traveller: a new look for PWR fresh fuel packages

    International Nuclear Information System (INIS)

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain

  16. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  17. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    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

  18. A new method for essential reduction of seismic and external loads on NPP's structures, systems and components

    International Nuclear Information System (INIS)

    A new method of essential decreasing of floor response spectra is proposed in the paper. The design considers interconnection of main structures inside NPP containment by High Viscous Dampers. In result within the same ZPGA level of a reference earthquake or the same intensity of other than seismic external loads the acceleration range of floor response spectra could be dropped in factor 2 to 3 or even more depending on a number of installed damping devices. That way an essential decreasing of dynamic loads on structures, components and systems (SSC) could be achieved. Another important advantage of the actual proposal consists of eliminating of possible interactions and collisions in the gaps between building structures that wears potential threat of building failure. In-depth dynamic analysis of the Reactor Building System with proposed design decision was undertaken and presented in the paper. Both a simplified stick model and accurate 3D Finite Element Model of PWR Reactor Building were investigated and comparative analysis results for the systems without dampers and with dampers were obtained. The proposed decision could be relatively simply implemented in design of a new Nuclear Plants as well as for existing one and upgrade its seismic, wind, blast and other kind of external event capacities. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  20. Selection of suitable sites for NPP in Slovenia (stage 3)

    International Nuclear Information System (INIS)

    Selection of suitable sites for nuclear power plants in Slovenia is considered. This includes the studies of available data on regional and local characteristics specified in general site suitability criteria for NPP. The most suitable selected sites will be included into land use urbanistic planning of Slovenia

  1. Software to Compare NPP HDF5 Data Files

    Science.gov (United States)

    Wiegand, Chiu P.; LeMoigne-Stewart, Jacqueline; Ruley, LaMont T.

    2013-01-01

    This software was developed for the NPOESS (National Polar-orbiting Operational Environmental Satellite System) Preparatory Project (NPP) Science Data Segment. The purpose of this software is to compare HDF5 (Hierarchical Data Format) files specific to NPP and report whether the HDF5 files are identical. If the HDF5 files are different, users have the option of printing out the list of differences in the HDF5 data files. The user provides paths to two directories containing a list of HDF5 files to compare. The tool would select matching HDF5 file names from the two directories and run the comparison on each file. The user can also select from three levels of detail. Level 0 is the basic level, which simply states whether the files match or not. Level 1 is the intermediate level, which lists the differences between the files. Level 2 lists all the details regarding the comparison, such as which objects were compared, and how and where they are different. The HDF5 tool is written specifically for the NPP project. As such, it ignores certain attributes (such as creation_date, creation_ time, etc.) in the HDF5 files. This is because even though two HDF5 files could represent exactly the same granule, if they are created at different times, the creation date and time would be different. This tool is smart enough to ignore differences that are not relevant to NPP users.

  2. NPP safety and personnel training. XII International conference. Abstracts

    International Nuclear Information System (INIS)

    The 12th International conference NPP Safety and Personnel Training took place in Obninsk, October 4-7, 2011. The issues of nuclear technologies safety are considered.The problems of life-cycle management of nuclear facilities are discussed. The criteria of assessment of physical protection systems of nuclear facilities are presented

  3. NPP Krsko core calculations to improve operational safety

    International Nuclear Information System (INIS)

    Calculation tools and methodology used to perform independent calculations of cumulative influence of different changes related to fuel and core operation of NPP Krsko were described. Some examples of steady state and transient results are used to illustrate potential improvements to understanding and reviewing plant safety. (author)

  4. Improvement of organization and payment on NPP construction

    International Nuclear Information System (INIS)

    Effect of introducing advanced methods and organization forms of work on increase of labour productivity during the NPP construction is shown. Role of progressive forms of payment for work in reducing terms and improving the quality of works performed is examined

  5. Progress in Decommissioning of Ignalina NPP Unit 1

    International Nuclear Information System (INIS)

    The aim of the paper is to present the Lithuanian legal framework regarding the nuclear safety in Decommissioning and Waste Management, and the progress in the Decommissioning Programme of the unit 1 of Ignalina Nuclear Power Plant (INPP). INPP is the only nuclear plant in Lithuania. It comprises two RBMK-1500 reactors. After Lithuania has restored its independence, responsibility for Ignalina NPP was transferred to the Republic of Lithuania. To ensure the control of the Nuclear Safety in Lithuania, The State Nuclear Power Safety Inspectorate (VATESI) was created on 18 October 1991, by a resolution of the Lithuanian Government. Significant work has been performed over the last decade, aiming at upgrading the safety level of the Ignalina NPP with reference to the International standards. On 5 October 1999 the Seimas (Parliament) adopted the National Energy Strategy: It has been decided that unit 1 of Ignalina NPP will be closed down before 2005, The conditions and precise final date of the decommissioning of Unit 2 will be stated in the updated National Energy strategy in 2004. On 20-21 June 2000, the International Donors' Conference for the Decommissioning of Ignalina NPP took place in Vilnius. More than 200 Millions Euro were pledged of which 165 M funded directly from the European Union's budget, as financial support to the Decommissioning projects. The Decommissioning Program encompasses legal, organizational, financial and technical means including the social and economical impacts in the region of Ignalina. The Program is financed from International Support Fund, State budget, National Decommissioning Fund of Ignalina NPP and other funds. Decommissioning of Ignalina NPP is subject to VATESI license according to the Law on Nuclear Energy. The Government established the licensing procedure in the so-called 'Procedure for licensing of Nuclear Activities'; and the document 'General Requirements for Decommissioning of the Ignalina NPP' has been issued by VATESI. A

  6. Polish media and public opinion on NPP Mochovce commissioning

    International Nuclear Information System (INIS)

    The so called 'Mochovce Problem' was one of the major topics in Polish media in the period from May to July 1998. The nuclear power plant commissioning caused an unexpectedly strong reaction, especially in the newspapers, slightly less so in electronic media. Faced with clearly hostile media reaction to Mochovce NPP, the National Atomic Energy Agency representatives, together with atomic and nuclear experts, undertook to change these attitudes. In numerous interviews, letters to the editors and talks with journalists, they attempted to correct the mistakes, explain the true safety situation in the nuclear power plant, by whom it was constructed, who supervised and tested the systems and so on. The completion of Mochovce NPP construction improved significantly the electricity balance in Slovakia, thus decreasing the pressure for continuing the operation of older Bohunice V1 units beyond their design lifetime. For this reason, as well as in view of striving for improvement in environmental factors beyond Polish southern border, especially after Kyoto/97 decisions on greenhouse gases emissions, the public opinion in Poland should support the Mochovce NPP construction. In 1996 Poland has signed with Slovakia a bilateral inter-governmental agreement on the prompt notification on nuclear accidents and on the cooperation in the nuclear safety and radiological protection matters. On the basis of this agreement the experts from Polish National Atomic Energy Agency are in perpetual contact with Slovakian Nuclear Regulatory Body and in each and every moment can obtain full and comprehensive information on the plant parameters iportant for nuclear safety. The experts explanations, together with the NAEA top management visit to the plant itself, brought some results. The media became less aggressive, and Polish public and authorities - contrary to the Austrians - do not protest loudly against the commissioning of this newest European NPP. Now, in December 1998, the tune of

  7. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    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

  8. Studies of a small PWR for onsite industrial power

    International Nuclear Information System (INIS)

    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

  9. Studies of a small PWR for onsite industrial power

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author)

  11. Experience and reliability of Framatome ANP's PWR and BWR fuel

    International Nuclear Information System (INIS)

    Based on three decades of fuel supply to 169 PWR and BWR plants on four continents, Framatome ANP has a very large database from operating experience feedback. The performance of Framatome PWR and BWR fuel is discussed for the period 1992-2001 with special emphasis on fuel failures, countermeasures and their effectiveness. While PWR fuel performance in most reactors has been good, the performance in some years did suffer from special circumstances that caused grid-to-rod fretting failures in few PWRs. After solving this problem, fuel of all types showed high reliability again. Especially the current PWR fuel products AFA 3G, HTP, Mark B and Mark BW showed a very good operating performance. Fuel reliability of Framatome ANP BWR fuel has been excellent over the last decade with average annual fuel rod failure rates under 1x10-5 since 1991. More than 40% of all BWR fuel failures in the 1992-2001 decade were caused by debris fretting. The debris problem has been remedied with the FUELGUARDTM lower tie plate, and by reactor operators' efforts to control the sources of debris. PCI, the main failure mechanism in former periods, affected only 10 rods. All of these rods had non-liner cladding. (author)

  12. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    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)

  13. Make use of EDF orientations in PWR fuel management

    International Nuclear Information System (INIS)

    The EDF experience acquired permits to allow the PWR fuel performances and to make use of better management. In this domain low progress can be given considerable financial profits. The industrial and commercial structures, the time constant of the fuel cycle, has for consequence that the electric utilities can take advantage only progressively of the expected profits

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

    Directory of Open Access Journals (Sweden)

    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.

  15. Long term Integrity of PWR Spent Fuel in Dry Storage

    International Nuclear Information System (INIS)

    The newly established organization KRMC (Korea radioactive waste management corporation) which is responsible for all kinds of radioactive waste generated in the Republic of Korea launched the PWR spent fuel dry storage research project in June 2009. This project has objectives to develop a storage system and evaluate the integrity of PWR fuel in dry storage. The project consists of three steps. At first step, it would develop own degradation models by referring to pre-exist good models and develop the hot test scenarios. Second step, test facilities would be constructed and used for testing the degradation behaviour in each mechanisms and in total. As a final step, total evaluation code would be developed by integrating each degradation model produced in the first step and the test data produced in the second step. All the activities would be summarized into a report and applied to licensing work. The Republic of Korea PWR spent fuels have unique characteristics of various fuel types (array type, clad material) and high capacity factor (maximum usage of fuel which is bad for integrity). These facts could impact on the research ranges of experimental data needed for degradation evaluation. In this research, spent fuel performance data concerning long term dry storage will be analysed and the major degradation mechanisms like creep and hydride behaviour will be studied and proposed for Korean PWR spent fuels

  16. A comparative study of fuel management in PWR reactors

    International Nuclear Information System (INIS)

    A study about fuel management in PWR reactors, where not only the conventional uranium cycle is considered, but also the thorium cycle as an alternative is presented. The final results are presented in terms of U3O8 demand and SWU and the approximate costs of the principal stages of the fuel cycle, comparing with the stardand cycle without recycling. (E.G.)

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

    Institute of Scientific and Technical Information of China (English)

    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.

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

    DEFF Research Database (Denmark)

    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...... and it is demonstrated that the levels of abstraction in models for diagnosis must reflect plant knowledge about goals and functions which is represented in functional modelling.Multi level flow modeling (MFM), which is a method for functional modeling,is introduced briefly and illustrated with a cooling system example...

  19. NPP Krsko Lifetime Extension - Business Impact for Hrvatska Elektroprivreda

    International Nuclear Information System (INIS)

    This paper deals with the analysis of possible business impacts for HEP in the case of NPP Krsko life extension. Due to numerous reasons nuclear power plant life extension of ten to twenty years is a common procedure abroad. Having this practise in mind as well as other circumstances in Croatian and Slovenian electric power system, the extension of NPP Krsko lifetime is considered to be a possible scenario. Foreseeable impacts of this decision are evaluated primarily with consideration of its effect on HEPs projected cash flows, though other aspects will be addressed as well. Preserving a well maintained production facility with an extraordinary operational record and stable, or possibly falling overall production costs seems as a very rational choice. This is particularly true having in mind expected rise of electricity demand and energy prices in the region. Having NPP Krsko in operation beyond 2023 implies that no replacement source for NPP Krsko capacity needs to be built. This means avoiding all costs connected with the construction and operation of the replacement plant, assuming it will be fossil fuelled. Due to the high uncertainty of the future fossil fuel prices, the avoidance of replacement plant operational cost is likely to prove as highly rewarding. It should be kept in mind that avoided costs also include the replacement plant greenhouse gases emission costs, thus further enlarging the list of value adding impacts. The latter is valid anticipating the ratification of the Kyoto protocol and joining the European emission trading scheme. In addition to that, the extension of NPP Krsko lifetime would mean that the majority of costs connected with the decommissioning and final waste disposal can be postponed further down the time line. This will have very positive financial and possibly technological impact. Other value creating effects for HEP that are foreseeable as a consequence of the plant lifetime extension include: maintaining the knowledge of

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

    Science.gov (United States)

    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

  1. Incorporation of the new safety technology in NPP design

    International Nuclear Information System (INIS)

    A permanent target of nuclear energy development is to optimize the quality of the NPP and the performance of the nuclear energy generation, but at the same time a remarkable effort was devoted to protect the public and the environment by means of the improvement of Nuclear safety principles. The incorporation of new technologies, safety criteria and Norms, analysis methods and concepts to enhance the design's safety of NPP as a result in some cases, of the interaction between operation and design work is a complex task in the case of NPPs in operation or under construction while the elaboration of advanced reactor concepts allows the simultaneous introduction of new safety features and modern technology. The purpose of this work is to summarize the Argentine experience in the field of nuclear safety improvement in our NPPs in operation/construction and in the new advanced reactor design. (author)

  2. The intelligent system for accident identification in NPP

    International Nuclear Information System (INIS)

    Accidental situations in NPP are of greet concern for operators, the facility, regulatory bodies and the environment. This work proposes a design of intelligent system aimed to assist the operator in the process of decision making when initiator events with higher relative contribution to the reactor core damage occur. The intelligent System uses the results of the pre operational Probabilistic Safety Assessment and the Thermal hydraulic Safety Analyses of the NPP Juragua as source for building its knowledge base. The nucleus of the system is presented as a design of an intelligent hybrid system from the combination of the artificial intelligence techniques: fussy logic and artificial neural networks. The system works with variables from the process of the firsts circuit, second circuit and the containment and it is presented as a model for the integration of safety analyses in the process of decision making by the operator when tackling with accidental situations

  3. Belene NPP - main approaches in the construction organisation

    International Nuclear Information System (INIS)

    Technical planning for the construction of two units at the Belene NPP is presented. Detailed scheme for the assembling of the containment is given, taking into account the capacity of cranes. For the construction of NPP Belene it is possible to use the existing on-site construction 50 tons cranes, as well as mobile crawler cranes with increased lifting capacity. A timetable for the construction is developed. A shortening of the construction time is possible with 4 months. Improvements of the construction process are suggested including use of ready pipe and equipment modules, automated welding, use of advanced technologies for isolation and concreting, organising and control of the technological processes, use of 3D and 6D modeling etc. The recommended improvements can shorten the construction period to 5-5.5 years

  4. Ageing of electronic equipment used in Cernavoda NPP

    International Nuclear Information System (INIS)

    This paper presents some ageing mechanisms and an evaluation of systems and electronic equipment from Instrumentation and Control (I and C) system, taking into account their importance in command, control and safety operation of NPP. Concepts regarding lifetime, single and multiple stresses, experimental evaluation methods of lifetime, evaluation norms, measurement methods and constraints of lifetime evaluation techniques are also presented. The analysis highlights the factors that influence the operational safety, improvements in the methods of design, manufacture and operation, typical causes of failures and a case study on ASFD modules (Amplifier for Flux Detectors Self-supplied). In order to create a data base regarding the behaviour in time of modules and equipment in similar conditions as those of Cernavoda NPP, measurements of the operational parameters, without input current and with injection of current, depending on the type of module designed for each type of detector self-supplied (with vanadium or platinum ) were carried out. (authors)

  5. Social problems and territorial experiences from closing down Tihange NPP

    International Nuclear Information System (INIS)

    The town of Huy, better known by the name of the NPP Tihange, which is the nearby village, accounts to 50,000 inhabitants including the surrounding region. An Advisory Committee at a community level was set up concerned with the emergency services and the NPP related problems. One of the outstanding problems is the planned interim storage for radioactive waste. The unsolved question of long term storage in Belgium resulted in the fact that general situation is not promising for launching the debate on the future of nuclear energy. The mission of the delegation from Huy visited Tokaimura with the aim to measure the scale of real responsibilities and the concern against nuclear power that the accident created in the population

  6. Appendix XXI: Workforce planning applications at the Paks NPP Ltd

    International Nuclear Information System (INIS)

    Paks Nuclear Power Plant Ltd., Hungary,is a state owned Public Stakeholders' Corporation with one nuclear site with four reactors in operation. The process describes the long term workforce planning policy at the Paks NPP Ltd. The objective of the human resource policy at the Paks NPP Ltd. is to make available personnel in optimal quantity, either employed or contracted, that are properly selected, trained and equipped with all knowledge, skills and attitudes required for the long term safe, competitive and reliable operation of the plant. Accordingly the primary goal achievable by the plant is job positions (requiring managerial, professional or special expertise) be filled in by staff members performing at the highest possible level of professionalism and the most efficient manner

  7. Geological-Hydrological Site Evaluation for NPP Planning

    Energy Technology Data Exchange (ETDEWEB)

    Faust, Brigitte; Mini, Paolo [Nordostsschweizerische Kraftwerke AG NOK, Parkstrasse 23, 5401 Baden (Switzerland)

    2008-07-01

    NOK is investigating the potential replacement of the current NPP in Beznau. In order to meet the requirements with respect to a general licence application, geological, seismological, and geotechnical engineering, but also hydrological boundary conditions have been defined. These conditions define the nature of necessary investigations to obtain the geological, seismic, geotechnical and hydrological data themselves forming the basis to determine the site suitability. Viability has to be provided that a NPP can be built and operated at the proposed site without compromising public health, safety and environment. The collected data are also the basis for the design of all safety relevant structures, systems, and components. For example, the latter have to withstand the effects of natural phenomena such as earthquakes and human induced impact such as airplane crash without loosing their capability to perform the assigned safety functions. (authors)

  8. Equipment performance monitoring in NPP Krsko (Summarized system health report)

    International Nuclear Information System (INIS)

    Management common goal is safe, reliable, effective, acceptable to public and conservative/cautious operation of NPP Krsko. A set of programs, including Corrective Action Program, Performance Indicators, Operating Experience, Self Assessment and System Health Report, is developed to assist NPP Krsko management in fulfilling those goals. System Health Report is a tool that management can use to quickly assess how selected systems are performing, to determine where additional management attention is required and to determine if appropriate corrective actions have been established. Summarized System Health Report is developed for management's quick overview of systems status, important system malfunctions and problems as well as major changes from previous assessment period. Summarized Report contains nine sections: status difference including brief explanation, selected performance indicators, new equipment problems, functional failures, important problem analyses, action plan for systems with Potential Danger (RED) status, maintenance rule status overview and systems availability (planned and unplanned).(author)

  9. Exposure management in emergency preparedness and planning at Mochovce NPP

    International Nuclear Information System (INIS)

    The legal requirements on emergency preparedness come from Acts of Slovak republic as the Act on State Safety Supervision of Nuclear Facilities, The Act on Civil Protection of population and the Act on Protection of Public Health. The operating organization, regulatory bodies and public authorities shall cooperate to prepare emergency plans according to these Acts in which the contents and conditions of the emergency planning are specified, exercises and reviews are demanded. On-site emergency plan with Emergency Plan Implementing procedures are the basic documents for emergency planning on NPP Management of response to emergency including exposure management shall be handled by Emergency Response Organization. There are the centers at which management is present during the emergency at NPP Mochovce: - Main Control Room; - Emergency Control Room; -Emergency Control Center; - Operational Support Center; - Technical Support Center; - Information Center; - Off-site Assessment Centre. NPP Mochovce owns facilities and equipment for the permanent environment monitoring at site and within the 20 km emergency planning zones. The major tasks of the Emergency Response Organization, which are carried out in compliance with defined procedures after an extraordinary event, are as follows: - assessment of extraordinary events; - warning and notification of the plant personal and population in the surroundings; - analysis and evaluation of on-site radiation situation; - exposure management; - taking measures for protection of the workforce (turn-back dose); - coordination of the activities of monitoring teams in environment; - coordination of the rescue teams (turn-back dose); - recommendations for public protection; - implementation of activities to remove the extraordinary event and to return to a safe state. Events are ranked in the Emergency Plan in compliance with their radiological consequences as follows: - 1. Degree - 'the alert state'; - 2.Degree; - 'On-site Emergency

  10. Acceptable risk as the criteria for NPP efficiency and safety

    International Nuclear Information System (INIS)

    Four types of criteria of effective and safe operation of NPP power units are studied and formulated. Criterion types are formulated in risk terms of failure to comply problems, losses of capital outlays and territory (ecological risk), NPP personnel and population health damage. The structure of losses and profiles under conditions of the modern economic system is considered for the given types of criteria. The mathematical cost-profit model of a stochastic processes of power unit utilization is developed. The explicit functional form of the positive effect on the process trajectories for each of criterion types is constructed. The functional construction is based on the representation of a power unit as a generator of random events leading to power unit failure its damage or loss of territory, correspondingly

  11. Application of reliability centered maintenance to Embalse NPP

    International Nuclear Information System (INIS)

    One of the most recent applications of Probabilistic Safety Analysis to Embalse NPP is the Safety Oriented Maintenance Program developed through the Reliability Centered Maintenance (RCM) methodology. Such an application was carried out by a cooperated effort between the staff of nuclear safety department of NPP and experts from Instituto Superior de Tecnologias y Ciencias Aplicadas of Cuba. So far 6 technological systems have been analyzed with important results regarding the optimization of preventive and predictive maintenance program of those systems. Any tasks of RCM were automated via MOSEG code. The results of this study were focused on the elaboration and modification of the Preventive Program, prioritization of stocks, reorientation of predictive techniques and modification in the time parameters of maintenance. (author)

  12. Unification as a method to increase NPP component quality

    International Nuclear Information System (INIS)

    The experience of unifjcation of standard sizes of applied materials (rolled stock, tubes) of structural elements (threads, radii, grooves), detail and equipment units in the course of reactors development is investigated. Examples of unification of the WWER-1000 reactor pipe connections, equipment units, NPP equipment unification level with WWER-440 and WWER-1000 reactors are indicated. Coefficients of equipment applicability which in author's opinion must be 30-70% are given

  13. Life extension program of KORI Unit 1 NPP in Korea

    International Nuclear Information System (INIS)

    The two phases of Life extension program for KORI Unit 1 NPP are presented. Phase I is completed. It was concluded that life extension is a feasible option in technical and economic aspects. Detailed analysis of RPV is underway, plan for Phase II is finished. It deals with screening and sorting of all relevant SSCs, detailed life evaluation of SSCs, ageing management program and documentation for license renewal application

  14. Surface Passivation Studies on n+pp+ Bifacial Solar Cell

    OpenAIRE

    Suhaila Sepeai; M. Y. Sulaiman; Kamaruzzaman Sopian; Saleem H. Zaidi

    2012-01-01

    Bifacial solar cell is a specially designed solar cell for the production of electricity from both sides of the solar cell. It is an active field of research to make photovoltaics (PV) more competitive by increasing its efficiency and lowering its costs. We developed an n+pp+ structure for the bifacial solar cell. The fabrication used phosphorus-oxy-trichloride (POCl3) diffusion to form the emitter and Al diffusion using conventional screen printing to produce the back surface field (BSF). Th...

  15. Full scale dynamic testing of Kozloduy NPP unit 5 structures

    International Nuclear Information System (INIS)

    As described in this report, the Kozloduy NPP western site has been subjected to low level earthquake-like ground shaking - through appropriately devised underground explosions - and the resulting dynamic response of the NPP reactor Unit 5 important structures appropriately measured and digitally recorded. In-situ free-field response was measured concurrently more than 100 m aside the main structures of interest. The collected experimental data provide reference information on the actual dynamic characteristics of the Kozloduy NPPs main structures, as well as give some useful indications on the dynamic soil-structure interaction effects for the case of low level excitation. Performing the present full-scale dynamic structural testing activities took advantage of the experience gained by ISMES during similar tests, lately performed in Italy and abroad (in particular, at the Paks NPP in 1994). The IAEA promoted dynamic testing of the Kozloduy NPP Unit 5 by means of pertinently designed buried explosion-induced ground motions which has provided a large amount of data on the dynamic structural response of its major structures. In the present report, the conducted investigation is described and the acquired digital data presented. A series of preliminary analyses were undertaken for examining in detail the ground excitation levels that were produced by these weak earthquake simulation experiments, as well as for inferring some structural characteristics and behaviour information from the collected data. These analyses ascertained the high quality of the collected digital data. Presumably due to soil-structure dynamic interaction effects, reduced excitation levels were observed at the reactor building foundation raft level with respect to the concurrent free-field ground motions. measured at a 140 m distance from the reactor building centre. Further more detailed and systematic analyses are worthwhile to be performed for extracting more complete information about the

  16. Nuclear Power Station Kalkar, 300 MWe Nuclear Prototype Power Station with Fast Sodium Cooled Reactor (SNR-300), Safety Report Reactor Core Mark-Ia

    International Nuclear Information System (INIS)

    The nuclear power station Kalkar is a prototype with a sodium cooled fast reactor (SNR-300) and a thermal power of 762 MW. The initial licensing procedure in 1972 was based on the so-called Mark-I core. During the following years, this core underwent some changes, for instance the thickness of the radial blanket was reduced to lower the electricity generation costs, the design of the absorber systems had been further optimized, and it became clear, that a full core with plutonium from MAGNOX-reactors could not be realized and that fuel from light-water reactors had also to be used. In this licensing document the modified reactor core Mark-Ia is described together with its assemblies and their loading procedure. The content of radioactive materials and the irradiation protection measures are discussed and those accidents are describe in an enveloping manner, from which an influence of the core modification cannot be excluded. Finally, both core versions (Mark-I and Mark-Ia) are compared with each other

  17. Commissioning of the THTR-300-MWe prototype power plant - A milestone for further application of this high-temperature reactor line

    International Nuclear Information System (INIS)

    With the completion of the THTR 300 and the development of the follow-on plant HTR 500, the BBC/HRB company group has taken the pebble bed high-temperature reactor to the threshold of the commercial stage. The HTR is an important innovation in the field of reactor technology which can play an important role in the intermediate and long-term supply of safe, environmental friendly and economic energy. The power level of 550 MW meets the requirements of the present energy market which shows a trend towards smaller power units as a result of grid size, investment effort, and the slower increase in electricity demand in industrial nations. The advantages of the high-temperature reactor, such as high thermal efficiency, low waste heat, low radiation exposure of operating and maintenance personnel, high inherent safety, simple mode of operation, flexible fuel cycle with the potential to extend fuel resources, high availability, are currently uncontested and will represent the future standards for the peaceful uses of nuclear energy. For special applications in industry (steam and electric power as a cogeneration product) and in case of special siting conditions (near industrial centers), BBC/HRB developed a small 100 MW HTR, which can also be constructed as a 200 MW twin plant at favorable cost conditions. For an economic use of domestic coal in a processed form, the HTR represents the optimum solution as to economic and environmental aspects as well as extension of resources, especially if combined with conventional gasification procedures and in direct application of nuclear process heat at high gas temperatures of about 950 deg. C. In this field the development of the heat-exchanging components remains to be completed, before commercial application will be possible. The HTR is particularly well suited for erection in developing countries and industrial threshold countries which turn to nuclear energy for the first time. On an international level the interest in the pebble bed high-temperature reactor has also increased recently. Thus the HTR is of great importance to electric power industry and industrial development

  18. Nuclear Power Station Kalkar, 300 MWe Nuclear Prototype Power Station with Fast Sodium Cooled Reactor (SNR-300), Short Description of the Reactor Core Mark-Ia

    International Nuclear Information System (INIS)

    The nuclear power station Kalkar is a prototype with a sodium cooled fast reactor (SNR-300) and a thermal power of 762 MW. The initial licensing procedure in 1972 was based on the so-called Mark-I core. During the following years, this core underwent some changes, for instance the thickness of the radial blanket was reduced to lower the electricity generation costs, the design of the absorber systems had been further optimized, and it became clear, that a full core with plutonium from MAGNOX-reactors could not be realized and that fuel from light-water reactors had also to be used. In this licensing document the modified reactor core Mark-Ia is described, and the radiological consequences of the core modification are quantified to be tolerable

  19. Integrated tool for NPP lifetime management in Spain

    Energy Technology Data Exchange (ETDEWEB)

    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)

  20. NPP Temelin safety analysis reports and PSA status

    International Nuclear Information System (INIS)

    To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

  1. Robot dispatching Scenario for Accident Condition Monitoring of NPP

    International Nuclear Information System (INIS)

    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

  2. Elements and strategies of Knowledge Management at Cernavoda NPP

    International Nuclear Information System (INIS)

    The paper highlights that there are Knowledge Management (KM) elements already implemented at Cernavoda NPP, ensuring the safe operation of the plant. The KM process is not formalized yet, but the following elements are part of existing KM practices for Cernavoda NPP organization: Plant policies and procedures are in place; Document control is in place using Document Management system; Configuration management is a complex and well structured process; Training and qualification is based on SAT; Learning from operating experience internal and external is a day by day practice, (procedures, PJB, PROEM, Simulator, RCA, Training, etc.); Work control system is implemented via Ventrix Work Management System; Corrective action tracking is implemented; Workforce planning and human resources management is a continuous process; Communications are implemented by the plant personnel; Company intranet and other web-based solutions (CBT) are put in place; Human performance improvement programme is on going. Also, it is reinforced the need to develop and implement a structured knowledge transfer and retention process for the Cernavoda NPP in order to preserve unique knowledge and skills that could be lost through attrition or planned staffing changes. (author)

  3. Enhancing NPP Safety Through an Effective Dependability Management

    International Nuclear Information System (INIS)

    Taking into account the importance of the continuous improvement of the performance and reliability of a NPP and practical measures to strengthen nuclear safety and security, it is to be noted that a good management for a nuclear power reactor involves a ''good dependability management'' of the activities, such as: Reliability, Availability, Maintainability (RAM) and maintenance support. In order to evaluate certain safety assessment criteria intended to be applied at the level of the nuclear reactor unit management, equipment dependability indicators and their impact over the availability and reactor safety have to be evaluated. Reactor equipment dependability indicators provide a quantitative indication of equipment RAM performances (Reliability, Availability and Maintenance). One of the important benefits of maintenance and failure data gathering is that it can be used as a support of probabilistic safety assessment (PSA). Also, a good dependability management implementation may be used to complement reactor level unit performance indicators in the field of safe operation, maintenance and improving operating parameters, as well as for Strengthening Safety and Improving Reliability of a NPP. This paper underlines the importance of nuclear safety and security as prerequisites for nuclear power. In addition, it demonstrates how different technical aspects, through implementation of a good dependability management, contribute to a strengthened safety and an improvement of availability of the NPP through dependability indicators determination and evaluation. (author)

  4. Feed water distribution pipe replacement at Loviisa NPP

    Energy Technology Data Exchange (ETDEWEB)

    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.).

  5. Corrosion Product Measurements to ensure integrity of the Steam Generators in Beznau NPP

    International Nuclear Information System (INIS)

    The Nuclear Power Plant Beznau comprises two identical 380 MWe PWR units with two loops each, commissioned in 1969 and 1971. Westinghouse was responsible for the primary part of the plant and BBC/ABB for the secondary circuit. The original materials used in the secondary systems were made of several copper-based alloys, such as for the Condensers, the Low Pressure Pre-heaters and the Moisture Separator Re-heater. The original Steam Generator Tubes were made of Inconel 600 MA. Regarding its age, the NPP Beznau has to be qualified as an old plant. However, in fact particularly in the last 20 years the plant has undergone an extensive modernisation programme in which about 1.5 billion Swiss Francs have been invested. Important measures were the replacements of the Steam Generators with tubes comprising Inconel 690 TT which was realized at unit 1 in 1993 and at unit 2 in 1999. Copper was completely banished from the secondary system and replaced by stainless and chromium steel. The Condensers were fitted with titanium tubes. The secondary water chemistry had to be changed by these replacements and moved step by step from Low-AVT with a pH of about 9.3 to High-AVT with a pH of 9.8 to 9.9, currently. To ensure the integrity of the new Steam Generators as well as of the whole Secondary System a corrosion product programme was introduced at the end of the Nineties. Several investigations which are performed periodically are represented by analyses of corrosion products, measurements of sludge mass and composition in the Steam Generators, Hide-Out-Return- and mass balance measurements of corrosion products in the whole circuit. Objectives of these investigations are assessments of the efficiency of the water chemistry and trend considerations regarding to the transport of corrosion products and pollutants into the Steam Generator, as well as of the potential danger of deposits and stored or absorbed pollutants. The main target of all measures is to avoid any chemical

  6. Technical and economical problems of decommissioning nuclear power plants (NPP) in Russia

    International Nuclear Information System (INIS)

    The introduction per new century has brought to atomic Engineering many new problems. One of them, which has got a serious urgency, we now shall consider. It is a problem of decommissioning NPP in Russia. By 2014 all maintained nowadays NPP in Russian Federation will develop the regular service life. And all of them on idea should be removed from operation. But, as we understand, in today's difficult economic conditions, to it NPP the procedure of prolongation of service life will be applied, and where it is impossible by virtue of the various reasons, the procedure of translation NPP in nuclear - safe condition and in a mode of a long storage under supervision, before acceptance of the decision about a method and way of financing of a decommissioning will be applied. Possible the following variants will be: use platforms of the old NPP for construction new NPP, or using as burial place NPP. The variant of a decommissioning up to a condition 'green grass' is represented unprofitable because of its dearness. The similar decommissioning was carried out in Japan. Was removed from operation research reactor of a type WWR. The expenses for this method of a decommissioning considerably surpass expenses for a method of a burial place NPP basically because of high cost of dismantle and transportation in long-term storehouses of the 1 contour equipment. The most urgent problem of decommissioning NPP, which developed their regular service life - is financing this final stage of a exploitation cycle of the block. I shall remind, that the financing is carried out from uniform fund of decommissioning. The formation of this fund occurs by deduction 1.3 % of cost of the put electric power to the consumers by all maintained NPP of Russia. The expenditure of this fund is carried on time on some tens years. They are spent for 3 basic stages: preparation to decommissioning NPP, long endurance under supervision, dismantle and burial the NPP equipment. Nowadays on faculty NPP MPEI

  7. Report on the PWR-radiation protection/ALARA Committee

    International Nuclear Information System (INIS)

    In 1992, representatives from several utilities with operational Pressurized Water Reactors (PWR) formed the PWR-Radiation Protection/ALARA Committee. The mission of the Committee is to facilitate open communications between member utilities relative to radiation protection and ALARA issues such that cost effective dose reduction and radiation protection measures may be instituted. While industry deregulation appears inevitable and inter-utility competition is on the rise, Committee members are fully committed to sharing both positive and negative experiences for the benefit of the health and safety of the radiation worker. Committee meetings provide current operational experiences through members providing Plant status reports, and information relative to programmatic improvements through member presentations and topic specific workshops. The most recent Committee workshop was facilitated to provide members with defined experiences that provide cost effective ALARA performance

  8. Industrial assessment of nonbackfittable PWR design modifications. Final report

    International Nuclear Information System (INIS)

    As part of the US Department of Energy's Advanced Reactor Design Study, various nonbackfittable PWR design modifications were evaluated to determine their potential for improved uranium utilization and commercial viability. Combustion Engineering, Inc. contributed to this effort through participation in the Battelle Pacific Northwest Laboratory industrial assessment of such design modifications. Seven modifications, including the use of higher primary system temperatures and pressures, rapid-frequent refueling, end-of-cycle stretchout, core periphery modifications, radial blankets, low power density cores, and small PWR assemblies, were evaluated with respect to uranium utilization, economics, technical and operational complexity, and several other subjective considerations. Rapid-frequent refueling was judged to have the highest potential although it would probably not be economical for the majority of reactors with the design assumptions used in this assessment

  9. Investigation, experiment and analysis on PWR sump screen clogging issue

    International Nuclear Information System (INIS)

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Flow clogging characteristics were investigated based on data for the relation of pressure loss and flow velocity during flow clogging due to debris accumulation. Deposition of chemical precipitates on the fuel cladding using an electrically heated rod was investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis with a thermal-hydraulic code on the downstream effect has shown that the core could be cooled because the core inlet flow compensates a evaporation of coolant due to the decay-heat even if core inlet was 99% clogged just after the ECCS recirculation operation started during the cold-leg break LOCA in PWR plants. (author)

  10. Investigation, experiment and analysis on PWR sump screen clogging issue

    International Nuclear Information System (INIS)

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the chemical effect and the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Chemical effect tests show that corrosion of carbon steel and galvanized steal may come to be important in domestic plants, in addition to corrosion of aluminum and insulator which has been considered dominant in the chemical effect. With respect to the downstream effect, deposition of chemical precipitates on the fuel cladding using an electrically heated rod is investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis on the downstream effect has shown that even if core inlet was completely clogged just after the recirculation operation started during LOCA in PWR plants, although upper part of core may be uncovered temporary and cladding temperature increased, core could be cooled by coolant injection through the hot-leg. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  12. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

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

    International Nuclear Information System (INIS)

    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)

  14. On catholyte application for hydrogen water chemistry in PWR

    International Nuclear Information System (INIS)

    Considering liquid water as a chemical compound with a wide band gap shows that its Redox potential as Fermi level in the band gap is the measurable characteristic of a non-stoichiometric aqueous coolant in recirculation system of PWR. The hypo-stoichiometric state with the negative Redox potential is realized when Fermi level is shifted to the bottom of conduction band. This state can be fixed by the electro-reduced water (catholyte) of the alkaline solution. Then, the hydride anions (H3O-) as proton acceptors and the hydrox-onium radicals (H3O) as electron donors are emerged in the alkaline catholyte and form hydrated clusters (AH)n(H2O)m of alkaline hydride. These particles as very strong reducers have a molar portion more than the gaseous hydrogen in the aqueous coolant and are the effective remedy for holding the negative Redox potential as an effect of hydrogen water chemistry in PWR. (authors)

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

    International Nuclear Information System (INIS)

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

  16. Three basic options for the management of PWR waste

    International Nuclear Information System (INIS)

    Relying on the national practices of France, Germany and Belgium, three reference management routes for PWR wastes were drawn up and subsequently evaluated in terms of costs and radiological impact. It was thus demonstrated that safety regulations and technical redundancies, especially for off-gas treatment, liquid waste processing and dry solid waste treatment, play an important part in the cost associated with each route. The analysis of the different treatment options for mixed solid low level waste highlighted the low cost effectiveness of incineration as compared to compaction. Whatever the scenario investigated, the disposal costs of PWR wastes proved to be quite marginal in the overall cost. The radiological impact associated with each route was assessed through individual doses resulting from liquid and gaseous effluents. This theoretical exercise included some sensitivity studies performed on a selection of important parameters

  17. PWR plant transient analyses using TRAC-PF1

    International Nuclear Information System (INIS)

    This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. ORNL-PWR BDHT analysis procedure: an overview

    International Nuclear Information System (INIS)

    The key computer programs currently used by the analysis procedure of the ORNL-PWR Blowdown Heat Transfer Separate Effects Program are overviewed with particular emphasis placed on their interrelationships. The major modeling and calculational programs, COBRA, ORINC, ORTCAL, PINSIM, and various versions of RELAP4, are summarized and placed into the perspective of the procedure. The supportive programs, REDPLT, ORCPLT, BDHTPLOT, OXREPT, and OTOCI, and their uses are described

  20. Improvement on main control room for Japanese PWR plants

    International Nuclear Information System (INIS)

    The main control room which is the information center of nuclear power plant has been continuously improved utilizing the state of the art ergonomics, a high performance computer, computer graphic technologies, etc. For the latest Japanese Pressurized Water Reactor (PWR) plant, the CRT monitoring system is applied as the major information source for facilitating operators' plant monitoring tasks. For an operating plant, enhancement of monitoring and logging functions has been made adopting a high performance computer

  1. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

    The behavior of a PWR fuel rod when operating in normal power cycles, excluding in case of accidents, is analysed. A computer code, that makes the mechanical analysis of the cladding using the finite element method was developed. The ramps and power cycles were simulated suposing the existence of cracks in pellets when the cladding-pellet interaction are done. As a result, an operation procedure of the fuel rod in power cycle is recommended. (E.G.)

  2. Propagation of nuclear data Uncertainties for PWR core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cabellos, O.; Castro, E.; Ahnert, C.; Holgado, C. [Dept. of Nuclear Engineering, Universidad Politecnica de Madrid, Madrid (Spain)

    2014-06-15

    An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for {sup 235,238}U, {sup 239}Pu, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.

  3. In-pile test of Qinshan PWR fuel bundle

    International Nuclear Information System (INIS)

    In-pile test of Qinshan Nuclear Power Plant PWR fuel bundle has been conducted in HWRR HTHP Test loop at CIAE. The test fuel bundle was irradiated to an average burnup of 25000 Mwd/tU. The authors describe the structure of (3 x 3-2) test fuel bundle, structure of irradiation rig, fuel fabrication, irradiation conditions, power and fuel burnup. Some comments on the in-pile performance for fuel bundle, fuel rod and irradiation rig were made

  4. Safety Analysis Report for the PWR Spent Fuel Canister

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Choi, Jong Won; Cho, Dong Keun; Chun, Kwan Sik; Lee, Jong Youl; Kim, Seong Ki; Kim, Seong Soo; Lee, Yang

    2005-11-15

    This report outlined the results of the safety assessment of the canisters for the PWR spent fuels which will be used in the KRS. All safety analyses including criticality and radiation shielding analyses, mechanical analyses, thermal analyses, and containment analyses were performed. The reference PWR spent fuels were in the 17x17 and determined to have 45,000 MWD/MTU burnup. The canister consists of copper outer shell and nodular cast iron inner structure with diameter of 102 cm and height of 483 cm. Criticality safety was checked for normal and abnormal conditions. It was assumed that the integrity of engineered barriers is preserved and saturated with water of 1.0g/cc for normal condition. For the abnormal condition container and bentonite was assumed to disappear, which allows the spent fuel to be surrounded by water with the most reactive condition. In radiation shielding analysis it was investigated that the absorbed dose at the surface of the canister met the safety limit. The structural analysis was conducted considering three load conditions, normal, extreme, and rock movement condition. Thermal analysis was carried out for the case that the canister with four PWR assemblies was deposited in the repository 500 meter below the surface with 40 m tunnel spacing and 6 m deposition hole spacing. The results of the safety assessment showed that the proposed KDC-1 canister met all the safety limits.

  5. Validation of gadolinium burnout using PWR benchmark specification

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Validation of gadolinium burnout using PWR benchmark specification

    International Nuclear Information System (INIS)

    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 Kinf, 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

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  9. QFLOOD-GT: a program for predicting PWR reflood

    International Nuclear Information System (INIS)

    A description is given of the present version of the QFLOOD-GT program for predicting the reflood stage of a large-break PWR loss-of-coolant accident. QFLOOD-GT has been developed from an earlier forced-reflood program which, using a conduction-controlled model for rewetting speed, gave good agreement with the FLECHT SEASET experiments. This earlier program has been incorporated into QFLOOD-GT as a subroutine called QFLOOD; in addition a downcomer model has been included in order to allow calculation of gravity reflood, and a computational scheme has been devised to simulate the chimney effect (the unequal distribution of inlet flow between hot and cool regions of the core). No quantitative comparisons between QFLOOD-GT predictions and integral-test data have yet been carried out, so the modelling decisions implemented in the program are at this stage unvalidated. Preliminary testing of the program has produced results which are for the most part qualitatively satisfactory. Calculations for indicative PWR conditions suggest that the chimney effect has a significant beneficial effect during PWR reflood, a conclusion in accordance with the findings of the Japanese 2D/3D experiments. (author)

  10. PWR experimental benchmark analysis using WIMSD and PRIDE codes

    International Nuclear Information System (INIS)

    Highlights: • PWR experimental benchmark calculations were performed using WIMSD and PRIDE codes. • Various models for lattice cell homogenization were used. • Multiplication factors, power distribution and reaction rates were studied. • The effect of cross section libraries on these parameters was analyzed. • The results were compared with experimental and reported results. - Abstract: The PWR experimental benchmark problem defined by ANS was analyzed using WIMSD and PRIDE codes. Different modeling methodologies were used to calculate the infinite and effective multiplication factors. Relative pin power distributions were calculated for infinite lattice and critical core configurations, while reaction ratios were calculated for infinite lattice only. The discrete ordinate method (DSN) and collision probability method (PERSEUS) were used in each calculation. Different WIMSD cross-section libraries based on ENDF/B-VI.8, ENDF/B-VII.0, IAEA, JEF-2.2, JEFF-3.1 and JENDL-3.2 nuclear data files were also employed in the analyses. Comparison was made with experimental data and other reported results in order to find a suitable strategy for PWR analysis

  11. Simulator training and human factor reliability in Kozloduy NPP, Bulgaria

    International Nuclear Information System (INIS)

    This is a PowerPoint presentation. Situated in North Bulgaria, in the vicinity of the town of Kozloduy, near the Danube River bank, there is the Bulgarian Kozloduy Nuclear Power plant operating four WWER-440 and two WWER-1000 units. Units 1 and 2 were commissioned in July, 1974 and November, 1975, respectively. These were shut down at the end of 2003. Units 3 and 4 were commissioned in December, 1980 and May, 1982. They were shut down at the end of 2006 as a precondition for Bulgaria's accession to the European Union. The 1000 MW units 5 and 6 of Kozloduy NPP were commissioned in September, 1988 and December, 1993, respectively. Large-scale modernization have been implemented and now the units meet all international safety standards. The paper describes the multifunctional simulator Kozloduy NPP for the operational staff training. The training stages are as follows: - Preparatory; -Theoretical studies; - Training at the Training Centre by means of technical devices; - Preparation and sitting for an exam before a Kozloduy NPP expert commission; - Simulator training ; - Preparation to obtain a permit for a license, corresponding to the position to begin work at the NPP; - Exams before the Nuclear Regulatory Agency (NRA) and licensing; - Shadow training at the working place; - Permission for unaided operation. The following positions are addressed by the simulator training: - Chief Plant Supervisor; - Shift Unit Supervisor; - Senior Reactor Operator; - Simulator Instructor; - Controller physicist; -Senior Turbine Operator; - Senior Operator of Turbine Feedwater Pumps of Kozloduy NPP. Improving of training method led to a reduction of number of significant events while worldwide practice proves that improvement of engineering resulted in an increase in the percentage of events, related to human factor. Analysis of human reliability in 2005 and 2006 in cooperation with representatives from Great Britain and the Technical University in Sofia were worked on the DTI NSP B

  12. Programmes design for Bohunice NPP personnel other than control room operators

    International Nuclear Information System (INIS)

    This paper deals with project development of training programmes for non-licenced NPP personnel-masters, field operators, maintenance and technical supporting personnel. The programme development focuses on the part stage and on the job training at NPP. Bohunice NPP belongs to plants with higher specific number of personnel per installed power capacity. This factor also influenced the choice of programmes design. Undermentioned procedure is one of various approaches to SAT exploitation for training programmes design. (author)

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

    Science.gov (United States)

    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.

  14. Attitude of the population and students to further construction of NPP

    International Nuclear Information System (INIS)

    This presentation shows the public opinion poll young people on NPP in the Slovakia i.e. comparative research results on Attitude of the Slovak population (1991 - a sample 1104 respondents, and Slovak students 1992 - a sample 291 respondents) to the nuclear energy and NPP; Source of information about NE and NPP, motivation of apprehensions of young people concerning NE; what meaning for young people have the arguments FOR and AGAINST construction of NPP in Slovakia; whom students from Bratislava believe during discussions about NE

  15. Total Spent Nuclear Fuel Back-End Costs For Ignalina NPP

    International Nuclear Information System (INIS)

    Discussions are going on different acceptable scenarios for Ignalina NPP decommissioning. To choose the optimum alternative it is necessary to know the spent nuclear fuel back-end costs for Ignalina NPP. This paper deals with an analysis of such results for different decommissioning scenarios. The costs are estimated for the main three components: NPP decommissioning and short-lived waste treatment costs, spent nuclear fuel (SNF) and long-lived waste (LLW) storage costs, SNF and LLW disposal costs. Also, the annual flow of these costs is shown The estimated data are compared with the available data on Ignalina NPP and Sweden. (author)

  16. PWR primary coolant sample lines - problems with measurement of corrosion products and experimental proposals for Ringhals PWR

    International Nuclear Information System (INIS)

    Coolant samples are drawn from PWR primary circuits through long narrow tubes. Concern that interaction with the sample line walls (by deposition and release) can result in inaccurate measurement of corrosion product concentrations has recently intensified after several observations of a dependence on sample line flow rate. Particularly significant instances of this have been observed at Ringhals PWR. A further problem is that measured concentrations show spurious transient increases after valving in the sample line. Sampling behaviour is complex since it involves particulate as well as soluble material, and deposition and release as well as localised phenomena associated with crud traps within the sample line. The present report has threefold function, firstly to review instances of anomalous sample line behaviour and secondly to present a basic theoretical background to aid interpretation of such behaviour. The third and most important function is to suggest plant measurements which might be made at Ringhals PWR to understand better the response of the sampling system by quantifying the effects due to corrosion product deposition on, and release from, sample line walls. (author)

  17. World's first NPP and its role in nuclear power development

    International Nuclear Information System (INIS)

    Full text: On 26th of June 2004, it will be 50 years anniversary of start-up of the World's First Nuclear Power Plant (FNPP). This small size nuclear power plant that was constructed on the site of Laboratory 'V' (now State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering) has become a symbol of peaceful use of nuclear energy. As long ago as before the Second World War and during the War, proposals on the use of nuclear energy for peaceful purposes including electricity production, were made by Academicians V.I. Vernadsky, S.I. Vavilov and P.L. Kapitsa. Based on these proposals, a Decree was issued by the Council of Ministers of the USSR in December 1946, in which the objectives of peaceful application of nuclear energy were first stated. However, target-oriented, intensive work on the designs of power reactors began after successful nuclear weapon test in 1949. Academician I.V. Kurchatov, manager of Nuclear Project of the USSR, became scientific supervisor of this work, while L.P. Beriya, Head of Special Committee and B.L. Vannikov, Head of the First General Board were administrative leaders. In May 1950, the Governmental Decree was issued on the construction at Laboratory 'V' of 'V-10' facility including three reactor units and separate machine building with a 5000 kW turbine generator. AM facility, the first nuclear power plant, was determined as the first priority. Initial evaluation of AM reactor neutronics was performed at Laboratory 2 under the leadership of I.V. Kurchatov, reactor design was developed at the SKB-5 NIIKHIMMASH under the leadership of Professor N.A. Dollezhal, and the LPI (A.I. Gutov) was the General Architect of the Project. On the proposal of I.V. Kurchatov, in March 1951, Laboratory 'V' became the scientific supervisor of justification of the First NPP physical characteristics. The First NPP was designed and constructed in four and a half years. Under condition of permanent lack of time and hard

  18. Use of niobium for accurate relative fast neutron fluence measurements at the pressure vessel in a WWER-440 NPP

    International Nuclear Information System (INIS)

    The reaction 93Nb (n,n') /sup 93m/Nb, with low threshold energy (approximately 0.9 MeV) was used in a novel way to obtain both absolute and accurate, relative experimental estimates for the fast neutron (threshold) fluence at the inner wall of the pressure vessel (PV) in a WWER-440 type PWR reactor. Niobium was separated using chemical methods from steel samples containing 0.5...1% niobium, which had been scraped from the inner wall of the Loviisa NPP. Irradiated niobium foils from the surveillance chains (SC) were also treated in a similar way. Niobium from the liquid samples was then electro-deposited in very thin layers on copper discs. The amount of niobium deposited was determined by spectrophotometry and liquid scintillation measurements. The accuracy of the mass determination was further checked by an additional and independent method. The absolute activity of the niobium deposits was determined by measuring the 16.6 and 18.6 keV X-ray emission with a calibrated Si(Li) detector. The threshold flux PHI/sub T/ (> 0.9 MeV) was then obtained. The results indicate that the relative fluence estimates (PV/SC), which are sufficient for relating the measured embrittlement of the SC steel samples to the embrittlement of the PV, can be obtained with many times reduced in-accuracy (from tens of per cent to less than ten per cent) as compared to separate absolute estimates and estimate obtained by normalizing the calculated fluxes to conform with the measured results at the SC. (author)

  19. Accident management insights after the Fukushima Daiichi NPP accident

    International Nuclear Information System (INIS)

    The Fukushima Daiichi nuclear power plant (NPP) accident, that took place on 11 March 2011, initiated a significant number of activities at the national and international levels to reassess the safety of existing NPPs, evaluate the sufficiency of technical means and administrative measures available for emergency response, and develop recommendations for increasing the robustness of NPPs to withstand extreme external events and beyond design basis accidents. The OECD Nuclear Energy Agency (NEA) is working closely with its member and partner countries to examine the causes of the accident and to identify lessons learnt with a view to the appropriate follow-up actions to be taken by the nuclear safety community. Accident management is a priority area of work for the NEA to address lessons being learnt from the accident at the Fukushima Daiichi NPP following the recommendations of Committee on Nuclear Regulatory Activities (CNRA), Committee on the Safety of Nuclear Installations (CSNI), and Committee on Radiation Protection and Public Health (CRPPH). Considering the importance of these issues, the CNRA authorised the formation of a task group on accident management (TGAM) in June 2012 to review the regulatory framework for accident management following the Fukushima Daiichi NPP accident. The task group was requested to assess the NEA member countries needs and challenges in light of the accident from a regulatory point of view. The general objectives of the TGAM review were to consider: - enhancements of on-site accident management procedures and guidelines based on lessons learnt from the Fukushima Daiichi NPP accident; - decision-making and guiding principles in emergency situations; - guidance for instrumentation, equipment and supplies for addressing long-term aspects of accident management; - guidance and implementation when taking extreme measures for accident management. The report is built on the existing bases for capabilities to respond to design basis

  20. Public acceptance of nuclear energy in Indonesia. A traditional puppet show for NPP in Java

    International Nuclear Information System (INIS)

    In Indonesia, activities on public relation have been carried out since the establishment of the agency responsible for research and development of atomic energy in 1958 (which later in the year 1964 became Batan, National Atomic Energy Agency), wile activities on public acceptance started 'at the end of 1985. After the Government indicated that it has to look seriously into the construction of the NPP, as from 1990, the public acceptance (PA) activities have been intensified in anticipation of the construction of the first NPP. An important part of the preparatory activities for the construction of the NPP is preparing the people, particularly those who live near the NPP site, to accept and to support the NPP program. The interdepartmental team has bee established in August 1990 consisting of representatives of several institutions, namely, Ministry of Information, Ministry of Domestic Affairs, Ministry of Defense, Land Utilization Agency, State Electricity Corporation, Office of the Minister for Population and Environment, and Batan. This team, called the 'Team on Public Acceptance of NPP', has the main task providing extensive information to the public concerning the government plan to construct nuclear power plants in Indonesia. From 1990 through the middle of 1994 the activity of the Team on Public Acceptance of NPP was focused on providing information on te NPP program to the local community leaders and the public, particularly in the districts of the NPP site in the Muria Peninsula region. Its main purpose was to precondition the people who live near the NPP site to accept the NPP in their neighbourhood without fear

  1. NPP/NPOESS Tools for Rapid Algorithm Updates

    Science.gov (United States)

    Route, G.; Grant, K. D.; Hughes, R.

    2010-12-01

    The National Oceanic and Atmospheric Administration (NOAA), Department of Defense (DoD), and National Aeronautics and Space Administration (NASA) are jointly acquiring the next-generation weather and environmental satellite system; the National Polar-orbiting Operational Environmental Satellite System (NPOESS). NPOESS replaces the current Polar-orbiting Operational Environmental Satellites (POES) managed by NOAA and the Defense Meteorological Satellite Program (DMSP) managed by the DoD. The NPOESS Preparatory Project (NPP) and NPOESS satellites will carry a suite of sensors that collect meteorological, oceanographic, climatological, and solar-geophysical observations of the earth, atmosphere, and space. The ground data processing segment for NPOESS is the Interface Data Processing Segment (IDPS), developed by Raytheon Intelligence and Information Systems. The IDPS processes both NPP and NPOESS satellite data to provide environmental data products to NOAA and DoD processing centers operated by the United States government. The Northrop Grumman Aerospace Systems (NGAS) Algorithms and Data Products (A&DP) organization is responsible for the algorithms that produce the Environmental Data Records (EDRs), including their quality aspects. As the Calibration and Validation (Cal/Val) activities move forward following both the NPP launch and subsequent NPOESS launches, rapid algorithm updates may be required. Raytheon and Northrop Grumman have developed tools and processes to enable changes to be evaluated, tested, and moved into the operational baseline in a rapid and efficient manner. This presentation will provide an overview of the tools available to the Cal/Val teams to ensure rapid and accurate assessment of algorithm changes, along with the processes in place to ensure baseline integrity.

  2. Training in fundamentals of radiological coverage in Laguna Verde NPP

    International Nuclear Information System (INIS)

    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)

  3. 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

    Institute of Scientific and Technical Information of China (English)

    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.

  4. NPP accident analysis for spray safety system operation

    International Nuclear Information System (INIS)

    Different methods for calculating the NPP containment spray system are considered in brief. The spray system consists of three manifolds mounted in the upper part of the containment. Indicated are formulas for calculating the degree of the liquid jet warm-up+ comparison of theoretical and experimental valwes of temperature and pressure in transients, dependence of pressure decrease rate on cooling water initial temperature, heat removal intensity curves at different cooling water temperatures and pressure decrease curves at various estimations of the degree of liquid jet warm-up

  5. About works perspectives in the ChNPP exception Zone

    International Nuclear Information System (INIS)

    The analysis of the situation that turned out by now in the 30-km exception zone of ChNPP has been carried out. The accumulation of the negative processes connected with the radionuclides transition to the soluble mobile forms are noted. The influence at least of three potentially dangerous high active sources on the increase of the subsoil water radioactive contamination level is determined. Recommendations concerning the organization of the reliable monitoring in the exception zone are given. The problems of the points of the radioactive waste provisional localization and collection are discussed. The top priority tasks connected with the management in the exception zone are defined

  6. Characterization of neutron field in a NPP workplace.

    Science.gov (United States)

    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.

  7. Vibrational fatigue analysis for main steam pipelines on a NPP

    International Nuclear Information System (INIS)

    The vibration and noise of pipings adjacent to isolation valves of main steam pipelines in a NPP were excessive during operation. To assess the safety in operation of main steam pipelines, both stress and fatigue analysis were performed using ANSYS program, and evaluation was performed according to ASME code. The main steam pipeline considered in this report, includes the main steam piping, safety valves, main steam isolation valve and other branch pipings between the containment penetration and the first transverse direction limit stop downstream of the isolation valve

  8. Application of ant colony optimization in NPP classification fault location

    International Nuclear Information System (INIS)

    Nuclear Power Plant is a highly complex structural system with high safety requirements. Fault location appears to be particularly important to enhance its safety. Ant Colony Optimization is a new type of optimization algorithm, which is used in the fault location and classification of nuclear power plants in this paper. Taking the main coolant system of the first loop as the study object, using VB6.0 programming technology, the NPP fault location system is designed, and is tested against the related data in the literature. Test results show that the ant colony optimization can be used in the accurate classification fault location in the nuclear power plants. (authors)

  9. Appendix XXVII: Succession planning applications at the Paks NPP Ltd

    International Nuclear Information System (INIS)

    Paks Nuclear Power Plant Ltd, Hungar, is a state owned Public Stakeholders' Corporation with one nuclear site with four reactors in operation. The process describes the complex career and succession planning policy at the Paks NPP Ltd. Making the competent labor force available is the responsibility of the corporate integrated HR system. A key item of this is operating the complex career and succession planning system, which covers the following prominent employee categories: - managerial posts; - non-managerial posts requiring higher academic education; - posts, that require medium level academic education and extensive professional experiences; - Posts of the operative scheme

  10. A risk informed safety classification for a Nordic NPP

    Energy Technology Data Exchange (ETDEWEB)

    Jaenkaelae, K. [Fortum Nuclear Services Ltd. (Finland)

    2002-01-01

    The report describes a study to develop a safety classification proposal or classi- fication recommendations based on risks for selected equipment of a nuclear power plant. The application plant in this work is Loviisa NPP unit 1. The safety classification proposals are to be considered as an exercise in this pilot study and do not necessarily represent final proposals in a real situation. Comparisons to original safety classifications and technical specifications were made. The study concludes that it is possible to change safety classes or safety signifi- cances as considered in technical specifications and in in-service-inspections into both directions without endangering the safety or even by improving the safety. (au)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. R and D Activities for NPP Life Management in Korea

    International Nuclear Information System (INIS)

    Many R and D activities related to PLiM have been carried out and implemented to Korean NPP for the long term operation. Those activities include ageing management study, periodic safety review, steam generator management programme, materials reliability programme, thinned pipe management programme, study on dissimilar metal welds, etc. Based on the R and D activities, long term operation of Kori Unit 1 was successfully started from January 17, 2008 for next 10 years beyond its original design life. In this paper, all the activities and their results of the R and D programmes are briefly introduced. (author)

  13. Comparison of the Uniform Hazard Spectra for Korean NPP Sites

    International Nuclear Information System (INIS)

    A site-specific median spectrum has been generally used for a seismic fragility analysis of structures and equipment in Korean nuclear power plants. The site-specific response spectrum, however, does not represent the same probability of an exceedance over the entire frequency range of interest. The uniform hazard spectrum is more appropriate to use in a seismic probabilistic risk assessment than the site-specific spectrum. In this study, the uniform hazard spectra were developed using the available seismic hazard data for Korean NPP sites

  14. Application of Robotic System for Emergency Response in NPP

    International Nuclear Information System (INIS)

    Increasing energy demand and concerns over climate change make increasing use of nuclear power plant in worldwide. Even though the probability of accident is greatly reduced, safety is the highest priority issue in the nuclear energy industry. Applying highly reliable and conservative 'defense in depth' concepts with the design and construction of NPP, there are very little possibilities with which accidents are occur and radioactive materials are released to environments in NPP. But NPP have prepared with the emergency response procedures and conduct exercises for post-accident circumstance according to the procedures. The application of robots for emergency response task for post-accident in nuclear facilities is not a new concept. Robots have been sent to recover the damaged reactor at Chernobyl where human workers could receive a lifetime dose of radiation in minutes. Based on NRC's TMI-2 Cleanup Program, several robots were built in the 1980s to help gather information and remove debris from a reactor at the Three Mile Island nuclear power plant that partially melted down in 1979. The first robot was lowered into the basement through a hatch and human operators monitoring in a control room drove it through mud, water and debris, capturing the initial post-accident images of the reactor's basement. It was used for several years equipped with various tools allowing it to scour surfaces, scoop samples and vacuum sludge. A second version carried a core sampler to determine the intensity and depth of the radiation that had permeated into the walls. To perform cleanup tasks, they built Workhorse that featured system redundancy and had a boom extendable to reach high places, but it was never used because it had too many complexities and to clean and fix. While remote robotics technology has proven to remove the human from the radioactive environment, it is also difficult to make it useful because it may requires skill about remote control and obtaining remote

  15. Appendix VIII: Knowledge management applications at the Paks NPP Ltd

    International Nuclear Information System (INIS)

    The process describes the knowledge management policy at the Paks NPP Ltd established to make available personnel in optimal quantity, properly selected, trained and equipped with all knowledge, skills and attitudes required for the long term safe, competitive and reliable operation of the plant with focus on the plant's service time extension recently approved. For this end - using knowledge related to the corporate recruitment needs and the job-specific requirements/expectations - personnel are selected from internal or external resources to become prepared through expedient, targeted development programmes to succeed in specific positions

  16. Problem of the NPP liquid radioactive wastes processing and disposal

    International Nuclear Information System (INIS)

    Modern methods of NPP radioactive waste processing and disposal are briefly presented, bituminization with the following disposal in the clayey soil in particular. Soviet installations of liquid waste bituminization and results of proving ground preservation of bituminic blocks are briefly described. These results indicate a possibility of bituminic material disposal with specific activity of 1 Ci/l directly in the soil without waterproofing. High safety and effectivity of waste bituminization is shown in comparison with preservation variant of liquid radioactive concentrates in capacities

  17. Waste management concept during dismantling of KKS NPP in Germany

    International Nuclear Information System (INIS)

    Full text: This paper gives an overview on the waste management of NPP Stade during dismantling. The general idea is to reduce the radioactive waste to about three percentage of the complete dismantling mass. The NPP Stade in Germany was shut down in November 2003. After a transient phase the license for dismantling was given in 2005. In the following 8 years about 20.000 tones of steel and 120.000 tones of concrete will be put out by the dismantling. The yearly output of steel will by about 100 times higher than during the running time of the NNP. For this a new processes for waste management had to be installed. The waste management during dismantling focus on free release (about 97%). Beside some minor exception, the rest is deemed to be radioactive waste. This will be collected in 1000 packages, which are ready to be sent to a final storage. As until now in Germany no final storage is open (and sending of radioactive waste to another country is forbidden), the NNP Stade build an intermediate storage, where the packages may by saved for longest 40 years. The clearance procedure in Germany is regulated in the radiation protection ordinances. It is based on a nuclide specific set of clearance levels. To fulfil these demands the NNP Stade chose a semi automated system for characterization and documentation, which we develop in accordance to our release license. It guaranties a most accurate determination of the relevant nuclides for a set of dismantling material (some 10 to 100 tones). After the characterization only the gamma-activity of the material is measured in boxes of about 500 kg. A short comparison of the chosen procedure with other options, possible in Germany will be given and the decision from the collaboration with the NPP in Barsebaeck, Sweden, will be withdrawn. Beside the free release different options are used for waste management, like incineration, sending to landfill or reuse in nuclear industry. The waste management of the NNP Stade take

  18. Ecological and environmental education in the region of Ignalina NPP

    International Nuclear Information System (INIS)

    During three last years the groups involving pupils, students and teachers take part in the educational sessions. They are organized in Drukshiai Ecological Station . The work of Ignalina NPP, its influence on the nature as well as methods and results of scientific investigation and assessment of the influence are discussed. To direct environmental research activity of scholars the special 3-8 days programs made up by researchers are used. They include theoretical and practical studies. Educational - research activity performed in the Drukshiai Ecological Station could be an example of ecological schooling center related to environmental protection. (author). 2 refs

  19. Characterization of neutron field in a NPP workplace

    International Nuclear Information System (INIS)

    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 Inst. 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. (authors)

  20. Sensitivity of Input Parameters in the PSHA for NPP Sites

    International Nuclear Information System (INIS)

    In this study, the most uncertain input parameter in the PSHA was identified for Korean NPP sites through sensitivity analysis. Shinuljin Nuclear Units 1 and 2 site was selected in this study. Expert panels developed input parameters for PSHA. Input data were analyzed, and used for sensitivity analysis to see the effects of each parameter on seismic hazard. Through sensitivity analysis, we identified the degree of uncertainty of each parameter. The CONPAS, EQHAZAS, and EQHAZAS Manager codes, which were developed by KAERI, were utilized to compute the seismic hazard. The Gutenberg-Richter parameter was identified as the most uncertain one in this study

  1. XIV International conference NPP safety and personnel training. Abstracts

    International Nuclear Information System (INIS)

    This volume presents papers of the participants of XIV International conference “NPP Safety and Training” held 25-27 November 2015 in Obninsk. Collection is prepared by teachers, research staff and students of NRNU MEPhI, INPE NRNU MEPhI, scientists and specialists of institutions of the State Corporation Rosatom, academic, scientific and industrial organizations in Russia and abroad. They reflect the level of researches in nuclear physics and energetics, nuclear medicine, electronics and automation, as well as advances in improving the efficiency and safety of nuclear power installations, environmental protection and training for the nuclear power industry

  2. Maintenance optimization of the Ignalina NPP safety systems

    International Nuclear Information System (INIS)

    This paper represents some theses and mathematical approach to optimization of technical specification problems of the diesel generators system (DGS) at the Ignalina NPP (INPP) in Lithuania. Analysis unites three chapters, each attributed to the solving of related problems: dependency of the Core Damage Frequency (CDF) on DGS test interval (Chapter 2), analysis and comparison of two different maintenance strategies for DGS (Chapter 3), DGS test interval's optimisation (averaged loss function minimisation) taking into consideration maintenance and DGS failure's consequence costs (Chapter 4). The mathematical model was developed which lets to estimate DGS failure probability, the averaged loss function value and study the influence of different parameters into final results. (author)

  3. Emergency operating procedures knowledge management for NPP MCR operators' support

    International Nuclear Information System (INIS)

    The described research aims at computerization of handling of the knowledge gathered from emergency operating procedures, which makes it possible to reduce frequency of errors committed by NPP main control room operators when executing the procedures. The mechanism for representation of procedures' declarative knowledge has been proposed. This method is based on two level semantic network hierarchy, namely intensional and extensional levels. The method has been used for a basic development of prototype of computer based procedures intended to support of operators under accidents. The system provides automatic on-line identification of the situation, selection of relevant procedure, its representation at video display unit and tracking of correctness of the procedure execution

  4. Dynamical study of the main building of the Paks NPP

    International Nuclear Information System (INIS)

    Explosion tests were carried out for determining the dynamical characteristics of the main building of the Paks NPP. The objectives of this work were on one hand the verification of the computations performed and, on the other hand we attempted measuring the damping value and evaluating the soil-structure interaction. The first series of explosions consisted of short-distance blows using 5- 40 kg of explosive, the second and third series were performed at longer distances with 50 - 500 kg loads. Explosions were performed by the Hungarian State geophysical Institute

  5. VVER reactor pressure vessels dosimetry for NPP lifetime management

    International Nuclear Information System (INIS)

    Assessment of the current state and lifetime of the Kozloduy NPP VVER-type reactors has been carried out by monitoring the irradiation exposure of reactor vessels and metal surveillance specimens. The neutron fluence determination is based on three-dimensional neutron transport calculation by the TORT code using the problem-oriented neutron cross-section library BGL, and by the MCNP code using the continuous energy dependence library DLC200. The fluence data are validated by measurements of the induced activity of threshold foil detectors. Reactor core loading patterns are determined to achieve minimal reactor vessel neutron exposure. (author)

  6. Characterization of neutron field in a NPP workplace.

    Science.gov (United States)

    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. PMID:17416593

  7. New approaches for seismic design standardization and retrofitting of NPP

    International Nuclear Information System (INIS)

    All the important structures, equipment piping of a nuclear power plant need to be safeguarded against natural disaster such as earthquake. The most effective way of reducing the earthquake loads in NPP structures is to isolate the structure using base isolators. Laminated rubber bearing, which is most commonly used as a isolation system for structures is discussed in the present paper. Passive energy dissipating devices called dampers are used for controlling the seismic loads in piping systems. Some of such dampers, lead extrusion dampers, friction dampers and elastoplastic dampers, are discussed in the present paper. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

    Energy Technology Data Exchange (ETDEWEB)

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

    DEFF Research Database (Denmark)

    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...

  11. Dose expenditure minimization during the NPP equipment and systems maintenance and repair

    International Nuclear Information System (INIS)

    Dose expenditure change depending on NPP operation term was estimated as well as maintenance and repair contribution to the collective dose depending on the reactor type. The ways to decrease the staff irradiation during the NPP maintenance were analyzed. The optimization relevance when radioactive protection choosing was shown

  12. REGULATORY FRAMEWORK AND EVALUATION OF HUMAN-MACHINE INTERFACES IMS NPP SAFETY CASE BASED METHODOLOGY

    OpenAIRE

    Харченко, В'ячеслав Сергійович; "Національний аерокосмічний університет ім.М.Є.Жуковського "ХАІ""; Орехова, Анастасія Олександрівна; "Національний аерокосмічний університет ім.М.Є.Жуковського "ХАІ""

    2012-01-01

    The problems associated with safety of human-machine interfaces, information and control systems in NPP are analyzed.. An approach to assess the safety HMI I&C system NPP, based on Safety Case methodology is proposed. The profile of standards for HMI quality requirements is presented. An example of HMI quality assessment is described.

  13. 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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  14. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  15. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

  16. Development of laser weld monitoring system for PWR space grid

    International Nuclear Information System (INIS)

    The laser welding monitoring system was developed to inspect PWR space grid welding for KNFC. The demands for this optical monitoring system were applied to Q.C. and process control in space grid welding. The thermal radiation signal from weld pool can be get the variation of weld pool size. The weld pool size and depth are verified by analyzed wavelength signals from weld pool. Applied this monitoring system in space grid weld, improved the weld productivity. (author). 4 refs., 5 tabs., 31 figs

  17. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author)

  18. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Sizewell B - analysis of British application of US PWR technology

    International Nuclear Information System (INIS)

    This report provides information on the staff's evaluation of major design differences and issues developed by the British in their application (Sizewell B) of US PWR technology. One design change, the addition of steam-driven charging pumps, was assessed to have a relatively high value compared to the other changes. However, the assessment is based on a number of assumptions for which inadequate data exist to make an unqualified judgment. Other changes to the US design (as typified by the SNUPPS design) were found to have relatively low or moderate safety benefits for US application

  20. Conversion ratio in epithermal PWR, in thorium and uranium cycle

    International Nuclear Information System (INIS)

    Results obtained for the conversion ratio in PWR reactors with close lattices, operating in thorium and uranium cycles, are presented. The study of those reactors is done in an unitary fuel cell of the lattices with several ratios V sub(M)/V sub(F), considering only the equilibrium cycles and adopting a non-spatial depletion calculation model, aiming to simulate mass flux of reactor heavy elements in the reactor. The neutronic analysis and the cross sections generation are done with Hammer computer code, with one critical apreciation about the application of this code in epithermal systems and with modifications introduced in the library of basic data. (E.G.)

  1. Axial simulation of PWR core and study of actuators

    International Nuclear Information System (INIS)

    Development of an operation code allowing to simulate the behaviour of a PWR type reactor core. Load following is controled by bore and control rods, taking into account the temperature counter-reactions. The fine behaviour of the fuel element during transients is not simulated, on the other hand the central part of the reactor is completely simulated. The regulation equation are easily modifiable and thus it is possible to test in open loop any modification brought about to this regulation. Description of simulation tests on CAS-2B reactor: core control, static tests, dynamic tests

  2. Modelling of pellet-cladding interaction in PWR's

    International Nuclear Information System (INIS)

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

  3. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    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)

  4. Highly Reliable NPP Instrumentation Using Constant Voltage Feedback Circuits

    International Nuclear Information System (INIS)

    A highly reliable nuclear power plant (NPP) instrumentation using constant voltage feedback circuits is proposed. Contrary to conventional NPP instrumentation, two operational amplifiers are used at auxiliary building to supply constant DC voltage across the potentiometer or wheatstone bridge type sensors, such as resistance temperature detectors (RTD) and strain gauges. The proposed constant voltage feedback circuits maintain its output voltage as constant regardless of the length of lead wire from the auxiliary building to the sensors. A detail analysis of the proposed feedback circuits and design procedures including the internal resistance and parasitic LC components of lead wire are presented. A prototype with lumped RLC values for modeling lead wires is fabricated and experimentally verified to supply constant 10V up to 200m distance under 0.8% error. Due to its versatile characteristics with cost effective structure, the proposed scheme can be generally extended to pressure meters and water-level recorders to guarantee robust measurements without conventional current transducers under severe accidents

  5. Performance indicator system with application to NPP management

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, J. [Iberdrola S.A., Madrid (Spain); Roldan, J. [Cofrentes NPP, Madrid (Spain)

    2001-07-01

    The objective of the paper is to present the work that is being conducted in the scope of a research project between Cofrentes NPP and the polytechnic university of Valencia aimed to the development and implementation of a performance indicators system to support plant management. In developing this system, attention is being paid to the areas of safety, production and dependability. The first step in the project was the development of the performance indicator system (PIS), in order to help in assessing the effectiveness of the different activities in plant (i.e. maintenance, inspections, tests, etc.). It is suggested establishing the operational indicators set in 3 levels. The lowest level concerns indicators monitoring performance and maintenance characteristics of components. The next one involves a subset of indicators placed at system level with a similar goal. And finally, the highest level summarizes the impact of the global policy in the whole plant from safety and performance point of view. The definition of an indicator should comprise, at least, the following items: indicator's name, performance area, definition and data needed. A strategy should define what, when and how indicators have to be evaluated, analyzed and reported. This article gives an example application of the methodology at the Cofrentes NPP, collective dose as safety indicator, power production as production indicator and the number of work orders as maintenance indicator are considered and their time evolution is given. (A.C.)

  6. Suomi NPP VIIRS Reflective Solar Bands Operational Calibration Reprocessing

    Directory of Open Access Journals (Sweden)

    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.

  7. Development of NPP Monitoring and Operation Support Technology

    International Nuclear Information System (INIS)

    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

  8. Development of NPP Monitoring and Operation Support Technology

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Modernization and power upgrading of the Loviisa NPP

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Review experience on overturning prevention of NPP buildings in Switzerland

    International Nuclear Information System (INIS)

    Buildings with high ratios of the building height to the width of their foundations show a great potential for overturning. This problem occurs under large horizontal actions like Safe Shutdown Earthquake (SSE) or Aircraft Impact (AI). There are different concepts to prevent overturning under these actions. During the design process one concept has to be chosen and investigated. Therefore, calculations need to be done, which can be static or dynamic and linear or non-linear. In most concepts the resistance and the activation potential of the foundation soil are important for the overturning stability. Therefore, the definition of nominal values and design values of the soil has to be done carefully. In most of the cases field testing is required due to non-existing data of the soil. Namely the transmission of tensile forces to the underlying soil or rock has to be supported by experimental verification of the system's behaviour. This paper shows in the first place different possibilities of overturning prevention, which were used at NPP buildings in Switzerland. In the second place a case study of a recently investigated NPP site is presented. Thereby, the potentials and the limitations of static and linear dynamic analysis are discussed and special experiences with non-linear dynamic analysis for earthquake and aircraft scenarios are presented. (authors)

  11. Improvements of Calculating of NPP Aging Management Program Costs

    International Nuclear Information System (INIS)

    NPP(nuclear power plant) AMP(aging management program) involves various costs to carry out plant refurbishments. PLiM(plant lifetime management) consists of many areas including economic evaluation of lifetime extension. In PLiM Phase 1 research project, with PLiMEE(Plant Lifetime Economic Evaluation) computer code we calculated and compared the costs of two cases which have the same assumption that they are providing the same output power for the electric grid. In addition to this assumption, we assumed that the preparation period for plant lie extension is 5 years and beside regular over haul days additional 20 days will be used in the plant refurbishment for that period each year. However, in phase 2, we are going to evaluate the residual life of SSCs(system, structure, components) and AMP will be recommended. In this situation, economic part of project should calculate AMP cost and benefit for plant life extension. In this paper, EPRI and Palisade economic computer codes are reviewed for this purpose, and the algorithms for calculating of NPP AMP costs are found out

  12. Integrated maintenance program at Cernavoda NPP - Unit 1

    International Nuclear Information System (INIS)

    The purpose of this paper is to provide the information related to the integrated maintenance program in the Cernavoda Candu plant. The integrated maintenance strategy is the one to provide guidance that can be used to enhance an existing maintenance program by improving the responsibility for plant components and systems maintenance while optimizing resources. The implementation of this guidance will help ensure that tasks for important components are applicable for the types of expected failures and are effective in controlling the failures. This document is showing how the maintenance program is controlled, integrated, enhanced and optimized in Cernavoda NPP. The content of the paper includes: A. The general description of the maintenance program including the main phases as follows: I. MAINTENANCE STRATEGY; II. MAINTENANCE PROGRAMS; Ill. MAINTENANCE PLANS; IV. INTEGRATED MAINTENANCE; V. MAINTENANCE EXECUTION; VI. MAINTENANCE OPTIMIZATION. B. Maintenance Program Objectives. C. Maintenance Performance Indicators. The paper also presents how technical assessment programs can be optimized and interfaced with the executions programs. It also includes a detailed one page description of the strategy plan (appendix one). The final conclusion of this paper is that only a complex comprehensive and integrated maintenance program as it was started and will be applied in NPP Cernavoda is assuring a safe and reliable production for the unit. (author)

  13. Possible satellite oceanography on coastal waters during the NPP stage

    Science.gov (United States)

    Zhu, J.; Asanuma, I.; Zhao, C.; Huang, B.

    2007-09-01

    Ocean color monitoring on the coastal water is still under study because of an incomplete atmospheric correction over the turbid water like over the coastal water along the China main land. Currently available sensors for science as MODIS on Terra or Aqua will terminate their service in the near future and the NPOESS Preparatory Project (NPP) will be the next satellite to support the satellite oceanography on the coastal water. The Tokyo University of Information Sciences (TUIS) has updated the MODIS receiving system to capture and ingest the Visible/Infrared Imager/Radiometer Suite (VIIRS) data from NPP, which will be launched in 2008. Data processing software from the Direct Readout Laboratory (DRL), such as the Real-time Software Telemetry Processing (RT-STPS), Simulcast, and DB algorithms, will be core programs in our system. VIIRS has seven bands in VIS&NIR, which are for ocean color research. The spatial resolution is 0.742×0.259 meters at nadir. While the MODIS spatial resolution of the nine ocean color bands is 1000m. The higher spatial resolution MODIS data (250 meters) is used to illustrate the advantage of the higher spatial resolution remote sensing data, such as data from VIIRS. In this study, we propose to combine the higher spatial resolution data with the traditional products of chlorophyll-a and sea surface temperature in the low resolution so as to extract further information on the coastal ocean.

  14. Status of Chinese NPP Industry and Nuclear Fuel Cycle Policy

    International Nuclear Information System (INIS)

    China still extended their experiences to both domestic and overseas so far. Chinese State Council approved its 'Medium and Long-term Nuclear Power Development Plan' in November 2007, indicating further definition for nuclear energy as indispensable energy option and future self-reliance development of nuclear industry. China intends to become self-sufficient not only in NPPs capacity, but also in the fuel production for all those plants. There are currently 17 NPPs in operation, and 28 NPPs under construction. However, domestic uranium mining supplying is currently less than a quarter of nuclear fuel demands. This paper investigated and summarized the updated status of NPP industry in China and Nuclear Fuel Cycle(NFC) policy. There still remain a number of technical innovation and comprehensive challenges for this nuclear developing country in the long-term, but its large ambitions and dramatic improvements toward future should not be ignored. As shown in this paper, the most suitable approach for China to achieve both environmentally-friendly power supplying and increasing energy demands meeting simultaneously must be considered. Nuclear energy now was recognized as the most potential and optimal way of energy supply system. In addition, to accommodate such a high-speed NPP construction in China, it should also focus on when and how spent nuclear fuel should be reprocessed. Finally, the nuclear back-end fuel cycle policy should be established, taking into accounts of all costs, uranium resource security, spent fuel management, proliferation resistance and environmental impact

  15. Readiness of the Mochovce NPP for unit 1 start-up with regards to meeting the requirements for increased safety

    International Nuclear Information System (INIS)

    A safety and reliability of operation of NPP type VVER 440/213, differences between Mochovce NPP and other VVER 440/213 NPPs contributing to the safety improvement, organization of the completion works of Mochovce NPP units 1 and technical condition of the installations are discussed

  16. Investigation of a possible emergency procedure for the VVER-1000 NPP in case of total loss of feedwater and a main steam line break

    Energy Technology Data Exchange (ETDEWEB)

    Muellner, N. [Institute for Risk Research, University of Vienna, Wien (Austria); Giannotti, W.; D' Auria, F. [DIMNP, University of Pisa, Pisa (Italy)

    2004-07-01

    Accident management procedures associated with nuclear power plant beyond design basis accidents should be developed with the aid of the more recent versions of advanced computational tools (best estimate codes) in order to verify the effectiveness of normal operation systems of the plant to avoid or minimize core damage. A. Madeira showed in her paper 'A PWR Recovery Option for a Total Loss of Feedwater Beyond Design Basis Scenario' that it is possible to safely control a total loss of feedwater scenario in the Angra-2 nuclear power plant, using two emergency procedures, namely the opening of the steam generator (SG) relief valves, and on the primary side the complete manual opening of all pressurizer relief and safety valves. This paper investigates the effectiveness of the procedure opening of the SG relief valves, followed by primary side feed and bleed for a generic VVER-1000 NPP in case of a total loss of feed water. The results indicate that the procedure is successful in reducing the primary side pressure and temperature to safe conditions, it means that long term core cooling is achievable. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  19. Degradation of fastener in reactor internal of PWR

    Energy Technology Data Exchange (ETDEWEB)

    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)

  20. Transient analysis of blowdown thrust force under PWR LOCA

    International Nuclear Information System (INIS)

    The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces obtained by Navier-Stokes momentum equation about a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a critical flow condition was satisfied. The following results are obtained. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break. (5) The blowdown thrust force in the analysis greatly depends on the selection of the exit pressure. (author)

  1. Degradation of fastener in reactor internal of PWR

    International Nuclear Information System (INIS)

    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)

  2. VERA Core Simulator Methodology for PWR Cycle Depletion

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Primary water chemistry for NPP with VVER-TOI

    International Nuclear Information System (INIS)

    Nowadays within the framework of development of the nuclear power industry in Russia the VVER-TOI reactor is under designing (Standard optimized design). The given design provides for improvement of operation safety level, of technical-economic, operational and load-follow characteristics, and for the raise of competitive capacity of reactor plant and NPP as a whole. In VVER-TOI reactor plant design the primary water chemistry has been improved considering operation experience of VVER reactor plants and a possibility of RP operation under load-follow modes from the viewpoint of meeting the following requirements: - suppression of generation of oxidizing radiolytic products under power operation; - assurance of corrosion resistance of structural materials of equipment and pipelines throughout the NPP design service life; - minimization of deposits on surfaces of the reactor core fuel rods and on heat exchange surface of steam generators; - minimization of accumulation of activated corrosion products; - minimization of the amount of radioactive processing waste. In meeting these requirements an important role is devoted to suppression of generation of oxidizing radiolytic products owing to accumulation of hydrogen in the primary coolant. At NPP with VVER-1000 reactor the ammonia-potassium water chemistry is used wherein the hydrogen accumulation is provided at the expense of ammonia proportioning. Usage of ammonia leads to generation of additional amount of radioactive processing waste and to increased irregularity of maintaining the water chemistry under the daily load-follow modes. In VVER TOI design the primary water chemistry is improved by replacing the proportioning of ammonia with the proportioning of gaseous hydrogen. Different process schemes were considered that provide for a possibility of hydrogen accumulation and maintaining owing to direct proportioning of gaseous hydrogen. The obtained results showed that transition to the potassium water chemistry

  5. Station Blackout Accident and Mitigation Measure for PWR-NPP%压水堆核电厂全厂断电事故及其缓解措施

    Institute of Scientific and Technical Information of China (English)

    张龙飞; 张大发; 徐金良

    2008-01-01

    以1座典型的3环路压水堆核电厂为参考对象,分别研究了发生全厂断电事故时堆芯在低压和高压状态下的损坏进程.结果表明:在考虑稳压器波动管的蠕变失效时,虽避免了高压熔堆,但低压状态下堆芯损坏更为严重,且产生更多的氢气.分析了导致这一结果的原因,提出了在堆芯出口温度达923 K时的严重事故缓解措施.计算结果表明:该缓解措施能有效地延缓堆芯损坏进程,为操纵员恢复交流电源以及采取其它缓解手段赢得更多时间.

  6. Interactive analysis of human error factors in NPP operation events

    International Nuclear Information System (INIS)

    Interactive of human error factors in NPP operation events were introduced, and 645 WANO operation event reports from 1999 to 2008 were analyzed, among which 432 were found relative to human errors. After classifying these errors with the Root Causes or Causal Factors, and then applying SPSS for correlation analysis,we concluded: (1) Personnel work practices are restricted by many factors. Forming a good personnel work practices is a systematic work which need supports in many aspects. (2)Verbal communications,personnel work practices, man-machine interface and written procedures and documents play great roles. They are four interaction factors which often come in bundle. If some improvements need to be made on one of them,synchronous measures are also necessary for the others.(3) Management direction and decision process, which are related to management,have a significant interaction with personnel factors. (authors)

  7. Dynamic analysis of the NPP pressure sensor system

    International Nuclear Information System (INIS)

    The aim of the work was to develop on-line methods of control and diagnostic of pressure sensors at a NPP. The analysis was carried out along two lines: i) The detection system including the sensor itself was modeled theoretically to obtain and study its transfer function, and to establish correspondence between the spectral characteristics of the pressure signal. ii) The numerical processing of the signal using the AR technique to reconstruct the transfer function and evaluate the system's response, to a step impulse, as well as the spectral analysis. The hydraulic model presented indicate that the spectral lines observed at ∼ 10 Hz are due to longitudinal oscillations of water in the system branches, while the response time of the detection system is effectively the time of signal retardation in the porous ceramic diaphragms of the δ-cell of the sensor itself. (author)

  8. Safe management of NPP ageing in the European Union

    International Nuclear Information System (INIS)

    The objectives of the study covered by this report consist of providing recommendations for the development of a methodology to monitor, control and anticipate the ageing of Nuclear Islands, in order to maintain their level of safety during the whole NPP life cycle. This is obtained through a synthesis of the potential ageing mechanisms, their potential effects and the available identification and mitigation methods, and an evaluation of the existing ageing management practices in Belgium, France and Spain, using international recommendations as guidance documents. The aspects covered include, in particular, classification of component priorities, identification of degradation phenomena, surveillance methods and preventive maintenance and repair / replacement programs. This synthesis comparison work is the basis for the recommendations issued at the end of the study. (author)

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

    Science.gov (United States)

    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

  10. Development of NPP personnel training system in Ukraine

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  12. The development of a living PSA for the Temelin NPP

    International Nuclear Information System (INIS)

    The general purpose of the project is to perform a systematic examination of unit 1 of the Temelin NPP for severe accident vulnerabilities by performance of a Level 1 and 2 Probabilistic Safety Assessment study. The objectives, scope of the work and project schedule are given. The analysis was focused on the occurrence of initiating events. The Event Sequence Diagram was used in accident sequence modelling. Fault trees were created using the NUPRA code. The approach used in human reliability analysis explicitly relates the operator actions to the working procedures and the success or failure of the system. A safety monitor was developed to provide an online measurement of risk based on the current plant configuration and testing status, thus enabling the plant staff to plan and perform maintenance activities meeting maximized safety and avoiding ionizer plant shutdown. (J.B.) 1 tab., 3 figs

  13. Risk management solutions for Cernavoda Unit #2 NPP completion

    International Nuclear Information System (INIS)

    The greatest risk facing today's electric utilities is change. Utility risk managers are being challenged to address the changes of deregulation, new technologies and changing work force. The utilities must be more aware of where all its costs are located to face with the challenge of competition, forcing them to respond with lower prices and innovative services. For completion of large projects, like Cernavoda NPP, Unit 2, the utilities are facing also with certain specific risks: politic, economic, social. The natural perils or machinery breakdown are common risks for operation as well as for construction-commissioning projects. Beyond the explicit challenges associated with the completion of a nuclear power plant in a transition economy environment, the utilities are facing with new risk factors such as professional liability, political risk, product warranty and liability, international exposure, etc. Changes in other governmental policies are also affecting large projects, mentioning healthcare, workplace safety, workers' compensation, environmental clean up, etc.(author)

  14. Neutron diffraction measurements of residual stresses in NPP construction materials

    International Nuclear Information System (INIS)

    Neutron diffraction is one of the most powerful methods for condensed matter studies. This method is used for non-destructive determination of residual stresses in material. The fundamental aspects of neutron diffraction are discussed, together with a brief description of the experimental facility. The principal advantage of using neutrons rather than the more conventional X-rays is the fact that neutron can penetrate deeply (2-4 cm for steel and more than 10 cm for aluminium) into metals to determine internal parameters within the bulk of materials. We present results of measurements residual stresses in NPP construction material - austenitic stainless steel (Cr-18%, Ni-10%, Ti-1%) coated with high-nickel alloy. (authors)

  15. Safe management of NPP ageing in European union

    International Nuclear Information System (INIS)

    The paper presents the results of a study conducted by Framatome, EDF, Belgatom, Unesa and IPSN, with the support of the European Commission in order to evaluate the practices and provide recommendations for the development of a methodology to monitor, control and anticipate the ageing of Nuclear Islands, in order to maintain their level of safety during the whole NPP life cycle. The study includes a synthesis of available European and International reports, the identification of potential ageing mechanisms, their effects and the available identification and mitigation methods, and an evaluation of the existing ageing management practices in Belgium, France and Spain, using international recommendations as guidance documents. Recommendations are proposed on regulatory aspects, management, prediction, detection, follow, mitigation of physical ageing, human management and adaptation of safety culture and procedures. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

  18. Far-field hydrogeological model of Paks NPP

    International Nuclear Information System (INIS)

    Complete text of publication follows. The settlements in the vicinity of Paks NPP gain the drinking water from the lower Pannonian aquifers. Uncontrolled releases from the NPP enter into the shallow ground-water and they are transported to the Danube in the Quaternary alluvial sandy gravel layer. The hydraulic potential in the area increases with the depth, so leakage from the upper gravel layer to the lower aquifer is not possible in undisturbed circumstances. The question is that an overexploitation of drinking water can change the pressure conditions so that the contaminated water can enter from the upper gravel layer into the lower aquifer. To answer this question the Visual MODFLOW program was used to simulate the groundwater w, providing quantitative estimates of w rates, w paths and travel times. The simulation was designed to represent average steady-state w conditions. A mathematical model is a representation of a real system or process. The subsurface environment constitutes a complex, three-dimensional, heterogeneous hydrogeologic setting, which strongly influences groundwater flow and transport, and it can be described through careful hydrogeologic practice in the field. The conceptual model of the hydrogeological system in the surroundings of Paks NPP was constructed from the available geologic, hydrogeological, geographical and meteorological data. To obtain a unique solution of a generic model corresponding to a given area, additional information about the physical state of the system is required (boundary and initial conditions). The model was constructed on a 20km x 22km area, its eastern boundary is Danube, the others are defined as no w boundaries. The model consists of 5 main layers: upper sequences of Pleistocene alluvial sediments (sandy layer, Kx,y = 10-4m/s); lower sequences of Pleistocene alluvial sediments (gravel, Kx,y = 10-3m/s); upper Pannonian sediments of restricted transmissivity (Kx,y = 5 x 10-6m/s); Pannonian aquifer (Kx,y = 10-4m

  19. Provisions of communication between NPP operational personnel: drifts of development

    International Nuclear Information System (INIS)

    Approaches to the creation of control room systems (CRS) in foreign countries are based on support of operator actions. The basis is the provision of compressed information about safety and technological processes with performance on screens of monitors and by paying special attention to human factors. In contrast to the traditional, the . advantage of such boards consists in granting to the operator a flexible format with the integrated information, with graphic, generalized, etc., representation of the data. When organizing information concerning the: operative personnel, the following principles are fixed: (a) Hierarchical performance of the information about NPP technological process and safety. (b) Provision of fast access to a required format or automatic performance of 8 required format in emergencies. (c) Hierarchical organization of the signal system; a uniform format of display organization. (d) Information support of the operative personnel at normal operation and in accident conditions. (authors)

  20. Preliminary study of the economics of enriching PWR fuel with a fusion hybrid reactor

    International Nuclear Information System (INIS)

    This study is a comparison of the economics of enriching uranium oxide for pressurized water reactor (PWR) power plant fuel using a fusion hybrid reactor versus the present isotopic enrichment process. The conclusion is that privately owned hybrid fusion reactors, which simultaneously produce electrical power and enrich fuel, are competitive with the gaseous diffusion enrichment process if spent PWR fuel rods are reenriched without refabrication. Analysis of irradiation damage effects should be performed to determine if the fuel rod cladding can withstand the additional irradiation in the hybrid and second PWR power cycle. The cost competitiveness shown by this initial study clearly justifies further investigations

  1. Maintaining staff competence-a NPP operator viewpoint

    International Nuclear Information System (INIS)

    For a nuclear power plant operator, it is crucial to guarantee the safe and economic operation of the power plant as well as to look after the general acceptability of nuclear power. As to human resources management, this requires continuous maintenance and enhancement of the performance of the individuals and organisation. To this end, several development projects have recently been implemented by Teollisuuden Voima Oy (TVO) at the Olkiluoto nuclear power plant, which consists of twin 840 MWe BWR units that commenced their operation in 1978 and 1980. Systematic initial and continuing training programmes are needed to maintain the technical and managerial skills and know-how at a high level. The present stabile state of nuclear power, i.e. operation of ageing plants with personnel ageing as well, requires a variety of actions to reinforce the training efforts. At Olkiluoto NPP, we have carried out an extensive modernization programme that allowed the personnel to strengthen their knowledge and supplement it with the most recent results of development. We have also closely monitored the NPP development projects of the vendors, which has added to the preservation of know-how and understanding of advanced nuclear power technology. We have close contacts to the research institutes and universities, and have performed R and D activities to limited extent. In addition to the projects mentioned above, a co-ordinated development programme, 'TVO 2002', was initiated last year. The main objective of this programme is to ensure the functional preconditions and the competitiveness of the company in a changing environment. The management and operational procedures will be developed in such a way that the goals set for year 2002 will be achieved. The programme is organised as ten projects, which cover a variety of development subjects. One of the focal areas includes projects that can be characterised with the words 'Survey of competencies' and 'Preservation of know

  2. Maintenance optimization methods for NPP critical systems, structures and components

    International Nuclear Information System (INIS)

    Considerable progress has been made in the development and implementation of the plant life management (PLiM) program with formal processes to identify systematically and evaluate the major critical systems, structures and components (CSSCs) in the station, and a plan to ensure that the plant surveillance, operation, and maintenance programs monitor and control component degradation well within the original design specifications essential for the plant life attainment. The paper presents a brief description of most important aspects of the methods used to identify the CSSCs taking into account maintenance activities. The system structures and components (SSCs) that influence decisively the NPP reliability are considered as critical. Also, for the accident conditions, the SSCs, which have a major influence to the system availability/operability, are considered also as critical. PSA techniques serve as a basis for maintenance optimization. In the paper a case of association between operating events and CSSC is presented. The main reference was the PSA study for Cernavoda NPP, Unit 1 CPSE B+. From this study it was selected and presented: - the contribution diagram for the occurrence of one accident (small LOCA) with some severe consequence (e.g. PDS 1 - 2 - early core damage); - dominant accident sequence analyzed; - the initial event that generated the possible dominant sequence; - the most important accident events; - qualitative and quantitative results; - main CSSC identification; - results validation and proposals for system operation performance improving; - documentation of the analysis. For the critical components, if periodical maintenance operations are needed, the maintenance activities must be optimized. In order to ensure these, some sensitivity studies have to be performed. In the system probabilistic models the associated events for the critical components could be the failure events of the components and also the unavailability due to component

  3. Use of PSA to prioritize NPP walk down activities

    International Nuclear Information System (INIS)

    Full text of publication follows: an international team of experts, organized by the Institute of Risk Research of the Academic Senate of the University of Vienna, conducted a walk down assessment of the Mochovce Nuclear Power Plant (NPP) in May 1998. The Mochovce NPP is a VVER-440/213 station which has been in final construction and implementation of a safety improvement program since 1996. In preparation for the walk down, available plant design information was organized into a probabilistic safety assessment (PSA) format in order to assist in prioritizing walk down activities. The PSA was also used to understand the likely impact on core damage frequency (CDF) resulting from implementation of the safety improvement program. The PSA employed event trees from the Dukovany PSA, modified to represent the system designs at Mochovce, to assess internal hazards. External hazards were assessed based on an evaluation performed in 1995, with consideration given to upgrades planned to be implemented up to 1998. Failure events were identified directly based on the Dukovany PSA and other sources, and quantified using data from the Dukovany PSA, other non-Mochovce PSAs, and generic data. The PSA was used to identify a suite of dominant accident sequences from internal and external causes. The impact of safety improvements on these sequences was then assessed. The 'pre-fix' and 'post-fix' cases were compared to identify the change in risk due to implementation of the safety improvement program. Following the walk down, the 'post-fix' case was updated based on insights gained from the walk down. This revised case was used to aid in understanding the risk significance of issues identified during the walk down. (author)

  4. Standard protocol for evaluation of environmental transfer factors around NPP sites

    International Nuclear Information System (INIS)

    This document presents the standard procedures for evaluation of site specific environmental transfer factors around NPP sites. The scope of this document is to provide standard protocol to be followed for evaluation of environmental transfer factors around NPP sites. The studies on transfer factors are being carried out at various NPP sites under DAE-BRNS projects for evaluation of site specific transfer factors for radionuclides released from power plants. This document contains a common methodology in terms of sampling, processing, measurements and analysis of elemental/radionuclides, while keeping the site specific requirements also in place. (author)

  5. IJS collaboration within the testing sequence for the NPP Krsko startup program

    International Nuclear Information System (INIS)

    Paper gives a description of work of a group of eight IJS members which have been working as NPP Krsko startup team under a contract between the NPP Krsko and IJS. Our work comprised activities during the core load in May 1981, hot functional testing, initial criticality, low power physical testing, initial synchronisation and 30%, 50%, 75% and 90% test sequences. Our collaboration in full power test sequence is planned for April. After this the testing program will be concluded and NPP Krsko should begin commercial operation. (author)

  6. Evaluation of Armenian NPP main structures dynamic characteristics according to experimental data

    International Nuclear Information System (INIS)

    It was envisaged and performed a complex observations for determination of main structures dynamic characteristics and assessment of ground-structure interaction while dynamic influences (explosions, micro oscillations), within the project of Seismic Conditions Additional Studies or Armenian NPP, in connection with NPP recommissioning Experimental factual materials obtained make a basis for coming to several interesting conclusions on dynamic characteristics of NPP main building and of turbo generators foundations, as well as on interaction level of ground-structure system. (author). 1 ref., 2 figs., 1 tab

  7. Evaluation of Armenian NPP main structures dynamic characteristics according to experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Arakelian, F.; Pogosian, H.; Pirouzian, S. [Armenergyseismicprojects Inst., Yerevan (Armenia)

    1995-12-31

    It was envisaged and performed a complex observations for determination of main structures dynamic characteristics and assessment of ground-structure interaction while dynamic influences (explosions, micro oscillations), within the project of Seismic Conditions Additional Studies or Armenian NPP, in connection with NPP recommissioning Experimental factual materials obtained make a basis for coming to several interesting conclusions on dynamic characteristics of NPP main building and of turbo generators foundations, as well as on interaction level of ground-structure system. (author). 1 ref., 2 figs., 1 tab.

  8. Qualitative and quantitative methods for human factor analysis and assessment in NPP. Investigations and results

    International Nuclear Information System (INIS)

    A description of the most frequently used approaches for human reliability assesment is given. The relation between different human factor causes for human induced events in Kozloduy NPP during the period 2000 - 2003 is discussed. A comparison between the contribution of the casual factors for events occurrences in Kozloduy NPP and an Japanese NPP is presented. It can be concluded that for both NPPs the most important casual factors are: 1) written procedures and documents; 2) man-machine interface; 3) environmental working conditions; 4) working practice; 5) training and qualification; 6) supervising methods

  9. User experiences using the operator support systems SCORPIO and SPDS at Kola NPP

    International Nuclear Information System (INIS)

    Full text: During the years from 2000 to 2004 a Safety Parameter Display System (SPDS) and a SCORPIO-VVER system have been installed and put into operation at Kola NPP in northern Russia. The projects have been carried out in co-operation with Finnish and Russian subcontractors with IFE, Halden, as the main contractor and the project manager. Both projects have been defined within the Norwegian Assistance Programme to Nuclear Installations at the Kola Peninsula and financed by the Norwegian government. The SPDS project has also been partly financed by the Finnish government. The Kola SPDS, which is based on a system originally developed for the Loviisa Nuclear Power Plant in Finland, was installed and put into operation on Kola NPP units 1 and 2 in April 2000 and on Kola NPP units 3 and 4 in March 2004. The system has been licensed and approved by Russian authority (GAN) for use in the control room environment. The SCORPIO-VVER system is based on the SCORPIO-VVER framework developed for Dukovany NPP, Czech Republic and Bohunice NPP, Slovak Republic. Further, the experience gained with the SCORPIO framework developed for Western PWRs have been utilised. The SCORPIO-VVER system was installed and put into operation at Kola NPP Nuclear Safety Centre in December 2003 for core monitoring on units 3 and 4. The user experiences by introducing the SPDS and the SCORPIO-VVER system at Kola NPP are so far very positive. By introducing an SPDS at Kola NPP, the process safety status (single or combinations of signals) is presented for the operator in a centralized location (on one screen) in the control room and in a form that makes it easier to obtain a quick assessment of the process safety status. According to feedback from control room personnel the SPDS has enhanced the monitoring of the safety status of the plant. The reactor core is the most central component of a nuclear power plant and a comprehensive core monitoring system is very important for maintaining reactor

  10. Role and responsibilities of management in NPP personnel training and competence. Working material

    International Nuclear Information System (INIS)

    The main aim and result of this seminar was imparting knowledge to various levels of Paks NPP management on their special tasks and responsibilities to achieve personnel competence, which include: meeting relevant regulatory and other requirements; defining the qualifications for NPP personnel jobs; training using systematic approach to training to attain the required level of qualification and competence of all NPP personnel, which includes management, operations, maintenance and technical support personnel and others; recruiting and retaining qualified personnel, including career development; supporting the training of all personnel on their responsibilities for introducing, maintaining and improving safety. Refs, figs and tabs

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

    DEFF Research Database (Denmark)

    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....

  12. Benchmark exercise on SBLOCA experiment of PWR PACTEL facility

    International Nuclear Information System (INIS)

    Highlights: • PWR PACTEL, the facility with EPR type steam generators, is introduced. • The focus of the benchmark was on the analyses of the SBLOCA test with PWR PACTEL. • System codes with several modeling approaches were utilized to analyze the test. • Proper consideration of heat and pressure losses improves simulation remarkably. - Abstract: The PWR PACTEL benchmark exercise was organized in Lappeenranta, Finland by Lappeenranta University of Technology. The benchmark consisted of two phases, i.e. a blind and an open calculation task. Seven organizations from the Czech Republic, Germany, Italy, Sweden and Finland participated in the benchmark exercise, and four system codes were utilized in the benchmark simulation tasks. Two workshops were organized for launching and concluding the benchmark, the latter of which involved presentations of the calculation results as well as discussions on the related modeling issues. The chosen experiment for the benchmark was a small break loss of coolant accident experiment which was performed to study the natural circulation behavior over a continuous range of primary side coolant inventories. For the blind calculation task, the detailed facility descriptions, the measured pressure and heat losses as well as the results of a short characterizing transient were provided. For the open calculation task part, the experiment results were released. According to the simulation results, the benchmark experiment was quite challenging to model. Several improvements were found and utilized especially for the open calculation case. The issues concerned model construction, heat and pressure losses impact, interpreting measured and calculated data, non-condensable gas effect, testing several condensation and CCFL correlations, sensitivity studies, as well as break modeling. There is a clear need for user guidelines or for a collection of best practices in modeling for every code. The benchmark offered a unique opportunity to test

  13. Frictional Behavior of Fe-based Cladding Candidates for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Hyung-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Byun, Thak Sang [Oak Ridge National Lab., Oak Ridge (United States)

    2014-10-15

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Knowledge-based diagnosis of PWR secondary water chemistry

    International Nuclear Information System (INIS)

    A prototype knowledge-based diagnostic system has been developed for more effective processing of the in-line chemistry sensor data from the PWR secondary water-steam circuit with the SUN 3/80 workstation and the Nexpert Object shell program. The system consists of the data interface, the data interpreter, the CHEMISTRY-expert, the ACTION-expert, and the user interface. The knowledge base defines physical and conceptual models of the target domain in a class/object hierarchy, giving rise to a reduced number of rules with pattern matching. The rule base is broken down into separate rule groups for task control, classification, prioritization, and diagnosis to minimize the inference time. The system is scheduled for the Verification and Validation test to collect operational information feedback in one of the Korea nuclear power plants in the near future. (author)

  16. Modeling local chemistry in PWR steam generator crevices

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  18. Failure probability of PWR reactor coolant loop piping

    International Nuclear Information System (INIS)

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria

  19. Development of high temperature adsorbent in PWR primary system

    International Nuclear Information System (INIS)

    Radiation exposure reduction in PWR is one of the most important problems to be solved. We have developed a high temperature Co adsorbent (HTA), which could be directly applied under primary reactor coolant conditions. This adsorbent was Fe-Ti-O system ceramics, and was fabricated to a suitable form for using in a packed column. Through those experiments of adsorption tests, compatibility tests, leaching tests and hot loop tests, it was found that HTA had superior adsorption capability to not only Co and Ni-ion but also many other transition metal ions. And it was also found that HTA was compatible with high temperature water, as well as advantageous for its waste solidification. Based on the experimental results, dose reduction effect was evaluated by a computer code. From this evaluation, it was found that more than 50 % dose reduction could be expected, when an advanced reactor coolant clean-up (RCC) system with HTA would be realized. (author)

  20. Sizewell 'B' PWR pre-construction safety report

    International Nuclear Information System (INIS)

    The Pre-Construction Safety Report (PCSR) for a PWR power station to be constructed as Sizewell 'B' is presented in 13 volumes containing 16 chapters. The PCSR has been submitted to the Nuclear Installations Inspectorate in support of the Central Electricity Generating Board's application for consent to the extension at Sizewell. It describes the design and provides the safety case for the proposed station, which comprises a 4-loop pressurized water reactor with associated generating plant and supporting auxiliary equipment. A general description of the station and its site is given. The strategy for ensuring nuclear safety is set out and the general design aspects of systems and plant outlined. The plant and systems, including their safety design bases and the fault analyses carried out for the design are described. Finally the way in which the plant will be decommissioned at the end of its useful life is outlined. (U.K.)

  1. Specification of water quality for the FRAMATOME PWR secondary circuit

    International Nuclear Information System (INIS)

    This paper describes the purpose, theory and scope of secondary system chemical specifications for FRAMATOME PWR nuclear power plants. All volatile treatment was chosen: controlling the feedwater pH by means of a volatile amine (ammonia, morpholine), and excluding oxygen by the addition of hydrazine. The pollutants are monitored at the steam generator drains by completely automatic measurements using simple and reliable techniques: pH measurement and a diagram of the cation conductivity versus sodium. An explanation is given of the monitoring techniques and to the effect of the various kinds of possible pollutant. A new concept is described, the annual quota expressed in day.microsiements.cm-1 which enables the amount of absorbed pollutants in the steam generator to be evaluated. The methods used for maintaining the desired chemical quality are dealt with

  2. Non linear identification applied to PWR steam generators

    International Nuclear Information System (INIS)

    For the precise industrial purpose of PWR nuclear power plant steam generator water level control, a natural method is developed where classical techniques seem not to be efficient enough. From this essentially non-linear practical problem, an input-output identification of dynamic systems is proposed. Through Homodynamic Systems, characterized by a regularity property which can be found in most industrial processes with balance set, state form realizations are built, which resolve the exact joining of local dynamic behaviors, in both discrete and continuous time cases, avoiding any load parameter. Specifically non-linear modelling analytical means, which have no influence on local joined behaviors, are also pointed out. Non-linear autoregressive realizations allow us to perform indirect adaptive control under constraint of an admissible given dynamic family

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Design of large steam turbines for PWR power stations

    International Nuclear Information System (INIS)

    The authors review the thermodynamic cycle requirements for use with pressurized-water reactors, outline the way thermal efficiency is maximized, and discuss the special nature of the wet-steam cycle associated with turbines for this type of reactor. Machine and cycle parameters are optimized to achieve high thermal efficiency, particular attention being given to arrangements for water separation and steam reheating and to provisions for feedwater heating. Principles and details of mechanical design are considered for a range both of full-speed turbines running at 3000 rev/min on 50 Hz systems and of half-speed turbines running at 1800 rev/min on 60 Hz systems. The importance of service experience with nuclear wet-stream turbines, and its relevance to the design of modern turbines for PWR applications, is discussed. (author)

  5. PWR pressurizer discharge piping system on-site testing

    International Nuclear Information System (INIS)

    Framatome PWR systems includes the installation of safety valves and relief valves wich permit the discharge of steam from the pressurizer to the pressurizer relief tank through discharge piping system. Water seal expulsion pluration then depends on valve stem lift dynamics which can vary according to water-stem interaction. In order to approaches the different phenomenons, it was decided to perform a test on a 900 MWe French plant, test wich objectives are: characterize the mechanical response of the discharge piping to validate a mechanical model; open one, two or several valves among the following: one safety valve and three pilot operated relief valves, at a time or sequentially and measure the discharge piping transient response, the support loads, the

  6. PWR-blowdown heat transfer separate effects program

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, D.G.

    1976-01-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results are obtained from the Thermal-Hydraulic Test Facility (THTF). Supporting experiments are carried out in several additional test loops - the Forced Convection Test Facility (FCTF), an air-water loop, a transient steam-water loop, and a low-temperature water mockup of the THTF heater rod bundle. The studies to date are described.

  7. CFD application to PWR subchannel void distribution benchmark

    International Nuclear Information System (INIS)

    A CFD study is performed to simulate the steady-state void distribution benchmark based on the NUPEC PWR Subchannel and Bundle Tests (PSBT). The CFD calculation predicted the void distributions in central typical and thimble subchannels, side subchannel and corner subchannel. The CFD prediction shows a higher void fraction near the heated wall and a migration of void in the subchannel gap region. A measured image of void distribution indicated a locally higher void fraction near the heated wall. The CFD predictions of void fraction and fluid density agree well with the measured ones for the low void test condition. However, the CFD calculations tend to underpredict the void fraction and overpredict the fluid density as the void fraction increases. (author)

  8. A system for trip analysis of PWR reactors using neural networks

    International Nuclear Information System (INIS)

    This work presents the basic concepts and the general description of a computational system developed for trip analysis in PWR nuclear power plants which is based on neural networks and artificial intelligence concepts. (author)

  9. N4 PWR makes full use of distributed processing and local networks

    Energy Technology Data Exchange (ETDEWEB)

    Aschenbrenner, J.F.; Tetreau, F.; Colling, J.M.

    1988-01-01

    The new instrumentation and control systems for the French N4 PWR power plant make extensive use of programmable controllers based on advanced microprocessor technology and distributed processing. Local networking techniques are widely used which simplify architecture and equipment design.

  10. Maintenance of Ni-based alloy at PWR plant

    International Nuclear Information System (INIS)

    Kansai Electric owns 11 PWR plants. At our PWR plants, we are taking various preventive maintenance measures on Ni-based alloy according to the prediction of possible trouble while past trouble occurred at overseas plants due to Primary Water Stress Corrosion Cracking (PWSCC) being considered. In addition, we are making an effort to put new maintenance techniques into practical use by conducting demonstration tests to confirm their applicability to actual plants. We have replaced reactor vessel heads at 7 plants with new ones. At the other 4 plants, we took, measures to reduce the temperature of reactor vessel head top to delay the timing of PWSCC occurrence. We are carrying out the constant load tests to predict the timing of PWSCC occurrence at these 4 plants. It is planned to conduct non-destructive inspections at an appropriate timing based on the result of the prediction. Based on the prediction of the timing of PWSCC occurrence at bottom-mounted instrumentation (BMI), we have developed water jet peening (WJP) technique to reduce residual stress and applied the technique to our plants successively. Meanwhile, a technique to cut and eliminate cracking has been developed. In addition, capping technique, which covers overall the concerned nozzle on the outer surface of the reactor vessel, has been also established. For alloy 132/82 weld metal for the connection, we are conducting ultrasonic inspection at our plants successively. In order to prepare against PWSCC occurrence, we have also established a technique to replace the entire section of concerned short piping with new one. (author)

  11. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    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.

  12. Westinghouse Passive Plants - AP600 and S PWR

    International Nuclear Information System (INIS)

    The original thought behind the AP600 passive design was that if the U. S. nuclear industry was to be revitalized, it would require a new, advanced technology with clearly proven benefits in safety. Response from the international arena indicates that, regardless of local domestic consideration, a revitalization of the U. S. industry is seen as very important, even essential, worldwide. And the potential for scale up of these passive safety features has been clearly established, allowing the benefits of the passive technology to be realized in countries that, for whatever reason, are interested in larger plant sizes only. Government projections indicate that U. S. energy demands in the 1990s will grow steadily, creating the need for approximately 117,000 to 322,000 MW of new generating capacity by the year 2010. Although this growth in electricity demand continues to be strong, orders for new nuclear power plants have not kept pace, in part due to licensing delays, prohibitive construction costs, and public uncertainty about safety. However, with the increased concerns about the environmental and economic security risks involved with an excessive dependence on fossil fuels, there is a growing realization that nuclear power must play a major role in our energy future. Looking to the future, Westinghouse is developing the AP600, a simplified two-loop PWR featuring passive safety systems. Drawing on the results of the AP600 development and testing programs, Westinghouse is also developing the larger S PWR, a passive, three-loop power plant with an output in the 900 to 1000 MW range

  13. First application of hollow fiber filter for PWR condensate polishing

    Energy Technology Data Exchange (ETDEWEB)

    Tsuda, S. [ORGANO Corp., Tokyo (Japan); Otoha, K.; Takiguchi, H. [Japan Atomic Power Co., Tokyo (Japan)

    2002-07-01

    In Tsuruga Unit-2 (PWR 1160 MWe commenced commercial operation in 1987), current procedure for secondary system clean-up before start-up had prolonged outage time and had consumed a huge amount of de-ionized (DI) water. In addition, iron oxide in condensate had accelerated the degradation of condensate demineralizer (CD) resin. The corrosion product of iron could also influence the secondary side corrosion of steam generator (SG) tubing if it intruded into SG through CD. To solve these problems, Japan Atomic Power Company (JAPC) decided to introduce hollow fiber filter (HFF) type condensate filter into Tsuruga-2, as the first application to PWR in the world. Because of retro-fitted HFF in Tsuruga Unit-2, limitations for installation space and flow resistance in condensate system and cost reduction required new design for compact and low differential pressure system and for long life filter module. JAPC and ORGANO assessed methodologies to achieve these goals. An advanced HFF system, including a newly developed compact HFF module design, was installed at Tsuruga Unit-2 in 1997 based on the assessment. During the 5 years since the installation, the HFF system has provided excellent crud removal that enables to shorten the outage period and to reduce DI water consumption drastically. Stable differential pressure (dP) trend of the HFF system indicates an expected module life of more than 7 years, with backwash cleaning required only 2 or 3 times per year. In addition to providing the expected operating cost reduction and improved SG tube integrity, numerous additional benefits have resulted from the retrofit. (authors)

  14. Identifying thermal cycling mechanisms in PWR branch line piping

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI, Charlotte, NC (United States); Keller, J.D.; Bilanin, A.J. [Continuum Dynamics, Inc., Ewing, NJ (United States)

    2002-07-01

    Predicting the onset and the characteristics of thermal cycling in pressurized water reactor (PWR) branch line piping systems is critical to formulation of thermal fatigue screening tools. The complex nature of the underlying thermal-hydraulic phenomena, however, significantly complicates prediction using analytical models or direct numerical simulations. Instead, it is necessary to perform scaled experiments to identify the physical mechanisms and to gather data for formulation of semi-empirical models for the thermal cycling phenomena. Through the EPRI Materials Reliability Program a test program is underway to identify and develop semi-empirical correlations for the physical thermalhydraulic mechanisms that cause thermal cycling in dead-ended PWR branch line piping systems. Three series of tests are being performed in this test program: configuration tests on a representative up-horizontal (UH) branch line piping geometry, configuration tests on a representative down-horizontal (DH) branch line piping geometry, and high Reynolds number tests to assess penetration of secondary flow structures into a dead-ended branch line. Results from UH and DH configuration tests indicate that random turbulence penetration is not sufficient for thermal cycling to occur. Rather a swirling flow structure, representative of a large, 'corkscrew' vortical structure, is required for thermal cycling. Scale tests on the UH configuration have simulated cycling phenomena observed in full-scale plant data and have been used to determine parametric sensitivities in formulating a predictive model for the thermal cycling. Data indicate that the mechanism for thermal cycling in UH configurations is stochastic but scales with the leak rate from the valve. The critical dependent variables are reduced to several non-dimensional scaling curves, resulting in a semiempirical predictive model. This paper discusses the test program and the results obtained to date. Application of these

  15. An Experience on RCS CRUD Sampling in European PWR Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Jae; Kim, Jong Bin [Sung Woo E and T Co., Seoul (Korea, Republic of); Kang, Duk Won [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    In most PWRs the normal method of corrosion product sampling is to collect a 'grab' sample from either the RCS hot or cold leg. This method is not ideal and the results are often dominated by soluble and particulate transients that can bias them high by factors of between ten and one hundred times. Nevertheless 'grab' samples can still give relatively satisfactory results from which qualitative trends of total soluble plus particulate corrosion product concentrations can be determined and, although 'grab' sampling may not be ideal, it is useful in detecting and following abnormal particulate releases from the core. It is possible to eliminate the worst of the transient effects by collecting a sample from continuously flowing RCS sample line, but the changes necessary to operate in this way are major, will be costly and may not be practicable for many existing plants. The evaluation of changes in corrosion product concentrations, particularly when the changes increase the particulate concentrations, can indicate that there is a risk that an Axial Offset Anomaly (AOA) may develop, or a risk of increased corrosion product releases when the plant shuts down for refueling. Recently, Diablo Canyon and Callaway in United States America, Ringhals in Sweden, Sizewell B in Great Britain, Vandellos in Spain and Doel in Belgium, these PWR plants have applied capillary sampling method to CRUD Analysis in parallel with grab sampling method under the recommendation of EPRI. In this thesis, it will show the practice based on actually tested method in European PWR plants.

  16. Estimating PWR fuel rod failures throughout a cycle

    International Nuclear Information System (INIS)

    A fuel performance engineer requires good prediction models for fuel conditions to help assure that any fuel repair operation he may recommend for the next refueling outage will have a minimal impact on nuclear plant operation. For nearly two decades, simple equilibrium equations have been used to provide estimates of the number of failed fuel rods in a pressurized water reactor (PWR) core. The unknown parameter is the isotopic escape rate (upsilon), which is often assumed to be --1 X 10/sup -8//s for the release of /sup 131/I from a 3- to 4-m-long PWR rod. The use of this escape rate value will generally produce end-of-cycle (EOC) predictions that are accurate within a factor of --3. When applied at the time when fuel rods initially fail, such as early in a reactor cycle, however, the prediction obtained may overestimate the number of failed rods present by a factor of 10 or more. While a goal of Combustion Engineering's (C-E's) efforts on failed fuel prediction (FFP) models over the past decade has been to increase the accuracy of the EOC estimate, recent efforts have emphasized improving prediction capability for failed rods present early in a reactor cycle. The C-E approach to modeling iodine release from failed fuel rods is based on dynamic escape rate theory that is incorporated in the C-E IODYNE (for iodine dynamic evaluation) code. This theory has been empirically modified to account for specific observed time dependencies of the release rates for /sup 131/I and /sup 133/I from a failed rod. In a current version of IODYNE, four such factors have been included in the FFP model, as described in this paper

  17. Materials ageing degradation programme in japan and proactive ageing management in NPP

    International Nuclear Information System (INIS)

    Predictive and preventive maintenance technologies are increasingly of importance for the long term operation (LTO) of Light Water Reactor (LWR) plants. In order for the realization LTO to be successful, it is essential that aging degradation phenomena should be properly managed by using adequate maintenance programs based on foreseeing the aging phenomena and evaluating their rates of development, where Nuclear Power Plants can be continued to operate beyond the original design life depending upon the regulatory authority rules. In combination with Periodic Safety Review (PSR) and adequate maintenance program, a plant life can be extended to 60 years or more. Plant Life Management (PLiM) is based upon various maintenance program as well as systematic safety review updated based upon the state of the art of science and technology. One of the potential life time limiting issue would be materials ageing degradation and therefore an extensive efforts have been paid world-widely. In 2007, NISA launched a national program on Enhanced Ageing Management Program and 4 nationwide clusters were formed to carry out the national program where materials ageing degradation was one of the major topics. In addition to these degradation modes, one important activities in this program is proactive materials degradation management directed by the author which is a kind of the extension program of NRC PMDA program based upon more fundamental approach by a systematic elicitation by the experts nominated from all over the world. NISA program can be divided into two phases, one is from fiscal years (FY) 2006 - 2010 and the other FY 2011. Later phase is focusing more on System Safety due to Fukushima NPP accident. The main objectives of the Phase I is to evaluate potential and complex degradation phenomena and their mechanisms in order to identify future risks of component aging in nuclear power plants. The following items are of particular concern in this phase: (a) investigation of

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Science.gov (United States)

    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.

  20. Joint probability safety assessment for NPP defense infrastructure against extreme external natural hazards

    Energy Technology Data Exchange (ETDEWEB)

    Guilin, L. [College of Engineering, Ocean Univ. ot China, Yushan Road No. 5, Qingdao 266003 (China); Defu, L. [Disaster Prevention Research Inst., Ocean Univ. ot China, Yushan Road No. 5, Qingdao 266003 (China); Huajun, L. [College of Engineering, Ocean Univ. ot China, Yushan Road No. 5, Qingdao 266003 (China); Fengqing, W. [Disaster Prevention Research Inst., Ocean Univ. of China, Songling Road No.238, Qingdao 266100 (China); Tao, Z. [College of Engineering, Ocean Univ. of China, Songling Road No.238, Qingdao 266100 (China)

    2012-07-01

    With the increasing tendency of natural hazards, the typhoon, hurricane and tropical Cyclone induced surge, wave, precipitation, flood and wind as extreme external loads menacing Nuclear Power Plants (NPP) in coastal and inland provinces of China. For all of planned, designed And constructed NPP the National Nuclear Safety Administration of China and IAEA recommended Probable Maximum Hurricane /Typhoon/(PMH/T), Probable Maximum Storm Surge (PMSS), Probable Maximum Flood (PMF), Design Basis Flood (DBF) as safety regulations for NPP defense infrastructures. This paper discusses the joint probability analysis of simultaneous occurrence typhoon induced extreme external hazards and compare with IAEA 2006-2009 recommended safety regulation design criteria for some NPP defense infrastructures along China coast. (authors)

  1. Safety improvement and results of commissioning of Mochovce NPP WWER 440/213

    International Nuclear Information System (INIS)

    Mochovce NPP is the last one of this kind and compared to its predecessors, it is characterized by several modifications which contribute to the improvement of the safety level. In addition based on Nuclear Regulatory Authority requirements and based on documents: - IAEA - Safety Issues and their ranking for NPP WWER 440/213, - IAEA - Safety Improvement of Mochovce NPP Project Review Mission, - Riskaudit - Evaluation of the Mochovce NPP Safety Improvements. Additional safety measures have been implemented before commissioning. The consortium EUCOM (FRAMATOME - SIEMENS), SKODA Praha, ENERGOPROJEKT Praha, Russian organizations and VUJE Trnava Nuclear Power Plants research institute were selected for design and implementation of the safety measures. The papers summarized, safety requirements, safety measures implemented, results of commissioning and results of safety analysis report evaluation. (author)

  2. Planning and management support for NPP personnel SAT-based training programmes

    International Nuclear Information System (INIS)

    Planning and management support for NPP personnel SAT-based training programmes is described for the following job positions: reactor operator; turbine operator; reactor maintenance worker; pump maintenance worker; chemistry foreman; health physics foreman; electric maintenance worker

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

    Data.gov (United States)

    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)...

  4. On geological and geophysical safety of the new state confinement for ChNPP-4

    International Nuclear Information System (INIS)

    The paper considers evidences and mechanisms of potential impact of local depression (microgeodynamic) zones on NPP safety. The depression forms were revealed within the Chornobyl NPP site, including location of the new safe confinement. Nature of zones and character of their influence is determined by fault structures and conjugated deep degassing processes, which cannot be detected and studied by traditional methods of engineering and exploration works during NPP siting. The assumption was made that significant and still unexplored risks for nuclear facilities are related to probable release of deep hydrogen through the bottom of depressions along sub-bottom channels. The approaches were proposed to studying the system of 'depression with its sub-bottom channel' within the sites of Chornobyl NPP and other nuclear power plants

  5. Information on current status of equipment and activities for extension of NPP Kozloduy life time

    International Nuclear Information System (INIS)

    The NPP Kozloduy personnel activities are described directed to undertake measures to overcome the difficulties entailed by equipment failures and human errors in order to avoid recurrent emergency events and to reduce their total numbers. 12 figs, 2 tabs

  6. Corrosion of copper-alloy tubes in the NPP cooling systems

    International Nuclear Information System (INIS)

    Factors efficiency on leaks in the NPP power unit steam turbine condensers are analyzed. Causes of corrosion - erosion wear of copper-alloy tubes are considered. A model for calculation of copper and copper-nickel alloy corrosion is proposed

  7. Joint probability safety assessment for NPP defense infrastructure against extreme external natural hazards

    International Nuclear Information System (INIS)

    With the increasing tendency of natural hazards, the typhoon, hurricane and tropical Cyclone induced surge, wave, precipitation, flood and wind as extreme external loads menacing Nuclear Power Plants (NPP) in coastal and inland provinces of China. For all of planned, designed And constructed NPP the National Nuclear Safety Administration of China and IAEA recommended Probable Maximum Hurricane /Typhoon/(PMH/T), Probable Maximum Storm Surge (PMSS), Probable Maximum Flood (PMF), Design Basis Flood (DBF) as safety regulations for NPP defense infrastructures. This paper discusses the joint probability analysis of simultaneous occurrence typhoon induced extreme external hazards and compare with IAEA 2006-2009 recommended safety regulation design criteria for some NPP defense infrastructures along China coast. (authors)

  8. Development of the Combination Method of PWR Spent Fuel for DUPIC Fuel Preparation

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Ju Ho; Kim, S. K.; Jung, T. C.; June, T. H.; Lee, J. M.; Kim, I. S.; Park, C. S.; Kim, M. J.; An, J. I.; Park, S. H. [Kyung Hee University, Seoul (Korea, Republic of)

    1997-07-15

    Optimum finding method of PWR spent fuel was developed in application of DUPIC fuel composition to nuclear fuel production. In order to make the database of the PWR spent fuel for the optimum composition, composition data of the PWR spent fuels from Youngkwang unit 1 and 2, Kori unit 3 and 4 and Uljin unit 1 and 2 were collected, analyzed and stored. Artificial intelligent access was attempted in optimizing the composition, and the combination algorithm for PWR spent fuel was developed. In this work database of the composition data of the PWR spent fuels from Youngkwang unit 1 and 2, Kori unit 3 and 4 and Uljin unit 1 and 2 as well as their combination algorithm for PWR spent fuel were developed. The combination algorithm is to find the combination of the spent fuel assembly which is quite close to the requirement per unit mass of DUPIC fuel. The required data are total weight of the fuel, tolerance of the errors, importance of the elements and the discharge data. This combination algorithm enables to find the optimum PWR spent fuel assembly for DUPIC fuel with the database of the spent fuels according to the DUPIC fuel standards. The combination algorithm developed in this work can afford the technical support to fuel supply in preparing the DUPIC fuel, and make contribution in DUPIC fuel cycle technology. It can be directly used in DUPIC fuel cycle technology, and can be also used in the management of the spent fuels with respect to their compositions and ingredients as well as the nuclear safeguards. Composition of the PWR spent fuel in each assembly depends on the initial concentration, degree of combustion, specific power, and its location in the reactor core. They may be affected by the kind of fuel rod and its axial length. Therefore, analysis procedures in these regards should be established for the effective application of the results of this work. 6 refs., 12 tabs., 46 figs. (author)

  9. Bias identification in PWR pressurizer instrumentation using the generalized liklihood-ratio technique

    International Nuclear Information System (INIS)

    A method for detecting and identifying biases in the pressure and level sensors of a pressurized water reactor (PWR) pressurizer is described. The generalized likelihood ratio (GLR) technique performs statistical tests on the innovations sequence of a Kalman filter state estimator and is capable of determining when a bias appears, in what sensor the bias exists, and estimating the bias magnitude. Simulation results using a second-order linear, discrete PWR pressurizer model demonstrate the capabilities of the GLR method

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Identification and evaluation of PWR in-vessel severe accident management strategies

    International Nuclear Information System (INIS)

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

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

    Institute of Scientific and Technical Information of China (English)

    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.

  13. The impact of the Ignalina NPP and Visaginas town on the economy of the region

    International Nuclear Information System (INIS)

    The present article is devoted to the influence of Visaginas town and Ignalina NPP on economy in the surrounding region. Preliminary the region includes the Ignalina and Zarasai administrative districts. However, these boundaries are relative. The analysis of the influence of Visaginas town and Ignalina NPP on the economy in the region revealed that such an influence exists, however, it is not as distinct as could have been expected from the town of such size. The influence on economy of the region includes influence on markets of economic resources (labour force and real estate markets were analyzed), influence on consumption market, influence on industries and services (the appearance of new manufacturing enterprises in the town itself, where, excluding the NPP, the economic sector is rather poorly developed, and the impact of the town on the enterprises of considered region were analyzed) and the influence on conditions for investment in the region. After the building Ignalina NPP and Visaginas the distribution of surrounding forests into protective categories has changed. The Ignalina NPP and Visaginas directly posses 1250 ha of forests. The influence of the town and Ignalina NPP on the enterprises of industry of considered region is important because Visaginas serving as a good market for various manufacturing enterprises makes it possible for them to maintain a certain economic level and have better profits. This aspect is extremely important for food and light industries. The extraction and building industry as well as many other branches of industries were most highly influenced during the building works of NPP. At present the building works meet only the current requirements of NPP and so they are not very intensive. The influence on consumption markets is more noticeable. With the building of Visaginas town consumption market of the region increased more than three times. The demand in the region increased as well as supply. The incomes of people in the

  14. Virtual reality in design, planning, operation and training related to the decommissioning of the Chernobyl NPP

    International Nuclear Information System (INIS)

    Virtual Reality in refueling operation and maintenance training at Leningrad NPP has been recognized by Chernobyl NPP. Institute for Energy Technology's is establishing now the Chernobyl Decommissioning Visualization Centre to be used for planning and training the dismantling procedures in addition to presenting it to the authorities and the public. It will be ready in 2007. The first scenario will be the dismantling of the refueling machines

  15. Ageing management and Long-term operation of NPP structures and components

    International Nuclear Information System (INIS)

    The paper presents methodological aspects for ageing management and long-term operation of NPP structures and components considering experience accumulated in nuclear area of Ukraine and IAEA recommendations. The research shows the role of ageing management for lifetime extension and justification of long-term operation of NPP structures and components. The given information is recommended to be used during development of regulatory and technical documents on ageing management and long-term operation

  16. Evaluation of radiation impacts of spent nuclear fuel storage (ISP-2) of Chernobyl NPP

    International Nuclear Information System (INIS)

    The estimation of radiation impacts during normal operation of dry spent fuel storage facility for ChNPP RBMK reactors, designed for long-term storage of spent fuel assemblies (SFA), which originated from the operation 1,2,and 3 units ChNPP was made. It is shown that the maximum concentration of radionuclides in the air near the ISF-2 well below allowable concentration for the population, and additional pollution is negligible in comparison with the existing one

  17. Underground NPP of mean and low power on the basis of ship nuclear technologies

    International Nuclear Information System (INIS)

    The principles of creating underground NPPs, melting higher requirements of nuclear and radiation safety, are described. The proposed NPP with the Pb-Bi-cooled fast neutron reactor is characterized by inherent self-protection, excluding origination of radiation-hazardous incidents. The experience of creation and operation of the reactors with the Pb-Bi coolant, accumulated by their application on submarines, may be used by developing such a NPP

  18. Analysis of quality control efficiency for NPP pipeline welded joints using nondestructive methods

    International Nuclear Information System (INIS)

    The problems of improving the quality of nondestructive examination for welded joints of NPP cooling system pipelines are considered. The data of ultrasonic testing and radiography of the welded joints of pipelines 800 mm in-diameter at the Smolensk and Kursk NPPs are analyzed. The conclusion on the necessity of accurate fulfilment of the existing PN AEh G-7-010-89 Rules of testing during NPP contruction and operation is made

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  20. A1 NPP Decommissioning, Slovakia. Annex A.I-6

    International Nuclear Information System (INIS)

    The first pilot NPP in the former Czechoslovakia was A1, which was built at Jaslovske Bohunice near the town of Trnava. An NPP with a capacity of 143 MW(e), it was commissioned in 1972 and operated with interruptions until 1977. A KS-150 reactor with natural uranium as fuel, D2O as moderator and gaseous CO2 as coolant was installed in the plant. The first serious accident associated with refuelling occurred in 1976, when a locking mechanism at a fuel assembly failed. The core was not damaged during that accident and after reconstruction of the damaged technology channel, the plant resumed operation. The second serious accident (level 4 according to the International Nuclear Event Scale) occurred in 1977, when a fuel assembly overheated, causing release of D2O into the gas cooling circuit. This accident was attributed to human error during replacement of a fuel assembly. Subsequent rapid humidity increase in the primary system resulted in damage to fuel elements in the core, and the primary system was contaminated by fission products. Internal reactor structures were also damaged. Radioactive contaminants penetrated into parts of the secondary system by leaking through steam generators. The radiation impact on and around the plant site was below specified limits for both events. Based on a technical and economic study of the difficult equipment repairs needed to restore plant operation, and also due to the policy decision to discontinue further construction of gas cooled reactors in the former Czechoslovakia, a decision was made in 1977 to terminate plant operation. The decision to proceed with the A1 plant decommissioning was issued in 1979. Beginning in 1981, decommissioning proceeded with disassembly of equipment from the secondary system (process equipment in the machine hall, turbines with auxiliaries, feed water tanks, diesel generator station, pumps, cooling towers, electric equipment). At the same time, other systems were disassembled, which included turbine

  1. Unusual occurrences during the whole operation of BN-250 NPP

    International Nuclear Information System (INIS)

    Unusual occurrences during the whole operation BN-350 NPP. 1. Oil ingress in high pressure receiver for the not reveled reason, 12.05.1994. 2. lncrease of water radioactivity of circulating water supply system due to heat exchanger leak of spent fuel assembly washing out system, 17.09.1993. 3. Lack of passableness of sodium drain header of primary circuit reveled during inspection on scheduled preventative maintenance, 28.11.1996. 4. Destruction of the blow-off line of MCP-6 due to corrosion damage of the pipeline while unit was being operated at rated power, 23.04.1993. 5. Lack of passableness of blow-down pipeline connecting reactor gas cover with gas-type pressurizer while unit was being operated at rated power, 17.11.1994. 6. Sodium ingress in blow-down pipeline of loop-5 intermediate heat exchanger while loop-5 was being fed of sodium during scheduled preventative maintenance, 27.06.1994. 7. Resistance deterioration of electro heating zones of loop-4 due to heat exchanger leak and water ingress in air-pipeline of primary circuit boxes recirculating air system, 02.05.1997. 8. Resistance deterioration of electro heating zones of sodium drain header of secondary circuit was sopped in the water for the extinguishing the fire of blowing ventilation oil-strainer, 23.12.1994. 9. Sodium ingress in gas-type pressurizer through pipeline of primary sodium cleanup system and blow-down pipeline of failed MCP-2 while primary sodium cleanup system was being connected to the primary circuit, 17.08.1976. As a rule, the main reactor systems are scrutinized more carefully than the auxiliary reactor systems and the order actions are existed for eliminating and mitigating of consequences of main reactor system fails. Therefore the auxiliary reactor system fails may impact on the main reactor systems through places of its contact in significant measure. The influence of auxiliary reactor system fails on main reactor systems and its possible consequences for behavior of the main

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  3. NPP simulator NSSS model validation the Krsko example

    International Nuclear Information System (INIS)

    NPP Krsko has obtained, as part of the plant modernization, a plant-specific full scope simulator with the capability to support in real-time, training for the complete range of operations which can be performed from the main control room and from other selected plant areas (remote shutdown panels, etc). Due to the requested approach the simulator was developed and validated in two distinct steps. First, it was validated for plant conditions in effect before SG replacement and power uprate (Cycle 15 and before); then a second step of validation was performed before and after initial startup in Cycle 17 (after SG replacement and power uprate). This approach gave NEK and CAE the opportunity to test extensively and validate the simulator based on real plant data as well as on the results of design basis accidents, i.e. transients. In this paper we focus on the validation of the NSSS model, based on real plant transients only. For comparison we also provide results obtained using more sophisticated thermohydraulic analysis tools, such as RELAP 5/mod2. (author)

  4. Derived Land Surface Emissivity From Suomi NPP CrIS

    Science.gov (United States)

    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.

  5. Using lessons learned to improve the NPP license renewal process

    International Nuclear Information System (INIS)

    There are 103 licensed, operating commercial nuclear power plant reactors in the United States today. Based on the Atomic Energy Act of 1954, the United States (U.S.) Nuclear Regulatory Commission (NRC) issues licenses for commercial nuclear power plants (NPP) to operate for up to 40 years and allows these licenses to be renewed. The NRC has established a license renewal process that establishes guidelines for renewal of operating power plant licenses for an additional 20 years of plant life and that can be completed in a reasonable period of time. The NRC has developed internationally available license renewal guidance documents to describe the interrelated aspects of preparing and reviewing license renewal applications. The objective of this presentation is to provide background information on the development of these documents and to briefly explain the intended use of the guidance documents singularly and in combination-to facilitate the renewal process starting from the application development by the plant to the regulatory staff review including audits and inspections of on-site supporting technical documentation. (authors)

  6. Analysis of seismic isolation problems of NPP buildings

    International Nuclear Information System (INIS)

    Conventional seismic design of earthquake-resistant structures is based on the philosophy conditions that the structural safety must be assured in extreme earthquake and the possible damage, if any, should be minimized or avoided in moderate earthquake ones. In the recent years the technique of seismic isolation appeared suitable to provide protection of buildings against seismic loadings introducing isolation devices (i.e. isolators) between the building itself and the foundation that may be adequate to shift the structures fundamental frequency outside the seismic characteristic range. The aim of this paper is to evaluate the influence of high damping rubber bearings, like the lead rubber ones, in the propagation of dynamic loading along NPP buildings. To this purpose a numerical approach was applied setting up suitable finite element models of the possible Generation III+ or IV NPPs. The influence of the isolator characteristics on the response of structures is analysed considering the variation of important parameters like the superstructure flexibility, isolation time period, etc. Several seismic analyses were therefore performed assuming as input different real earthquake ground motions as available in literature and in the relevant rules. The obtained results were herein presented and discussed in order to highlight the isolation technique effectiveness. (authors)

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

    Directory of Open Access Journals (Sweden)

    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.

  8. Safety analyses of the electrical systems on VVER NPP

    International Nuclear Information System (INIS)

    Energoprojekt Praha has been the main entity responsible for the section on 'Electrical Systems' in the safety reports of the Temelin, Dukovany and Mochovce nuclear power plants. The section comprises 2 main chapters, viz. Offsite Power System (issues of electrical energy production in main generators and the link to the offsite transmission grid) and Onsite Power Systems (AC and DC auxiliary system, both normal and safety related). In the chapter on the off-site system, attention is paid to the analysis of transmission capacity of the 400 kV lines, analysis of transient stability, multiple fault analyses, and probabilistic analyses of the grid and NPP power system reliability. In the chapter on the on-site system, attention is paid to the power balances of the electrical sources and switchboards set for various operational and accident modes, checks of loading and function of service and backup sources, short circuit current calculations, analyses of electrical protections, and analyses of the function and sizing of emergency sources (DG sets and UPS systems). (P.A.)

  9. Suomi NPP VIIRS Striping Analysis using Radiative Transfer Model Calculations

    Science.gov (United States)

    Wang, Z.; Cao, C.

    2015-12-01

    Modern satellite radiometers such as VIIRS have many detectors with slightly different relative spectral response (RSR). These differences can introduce artifacts such as striping in the imagery. In recent studies we have analyzed the striping pattern related to the detector level RSR difference in VIIRS Thermal Emissive Bands (TEB) M15 and M16, which includes line-by-line radiative transfer model (LBLRTM) detector level response study and onboard detector stability evaluation using the solar diffuser. Now we extend these analysis to the Reflective Solar Bands (RSB) using MODTRAN atmospheric radiative transfer model (RTM) for detector level radiance simulation. Previous studies analyzed the striping pattern in the images of VIIRS ocean color and reflectance in RSB, further studies about the root cause for striping are still needed. In this study, we will use the MODTRAN model at spectral resolution of 1 cm^-1 under different atmospheric conditions for VIIRS RSB, for example band M1 centered at 410nm which is used for Ocean Color product retrieval. The impact of detector level RSR difference, atmospheric dependency, and solar geometry on the striping in VIIRS SDR imagery will be investigated. The cumulative histogram method used successfully for the TEB striping analysis will be used to quantify the striping. These analysis help S-NPP and J1 to better understand the root cause for VIIRS image artifacts and reduce the uncertainties in geophysical retrievals to meet the user needs.

  10. Peculiarities of soil mycobiota composition in Chernobyl NPP

    International Nuclear Information System (INIS)

    Soil mycobiota in zones of strong radiation control (Lyutizh region), nine stationaries in 30 km Chernobyl NPP zone and one in the vicinity of Kiev has been investigated. 180 species of 70 genera of micromycetes has been isolated and identified. The species of genera Penicillium, Mortierella, Trichoderma, Cladosporium dominated in all soils studied, and the species of genera Acremonium, Paecilomyces, Oidiodendron etc. were frequent. The comparison of the generic composition of micromycetes in localities studied has revealed their similarity. The characteristic feature for the mycobiota was regular revealing and specific variability of melanin-containing fungi. Many of them compose the reserve group of micromycetes. There was found new and rare species Acrodontium crateriforme (v.Beyma) de Hoog, Botryosporium longibrachiatum Oud., Chuppia sarcinifera Deighton, Harzia acremonioides Harz etc., with were not registered in these soils before the accident. The structure of the micromycete complexes presented as a correlation Pleiades has been established. Discovered Pleiades were formed mainly by melanin-containing fungi. According to the preliminary results the micro-mycetes complexes, taking into account their occurrence, can be used as indicators of unfavourable ecological soil state. To our mind, a search for such complexes may be conducted among of melanin-containing species as well as numerous species of genus Penecillium and related Paecilomyces

  11. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  12. Planning and management support for NPP personnel SAT-based training programmes. IAEA regional workshop. Working material

    International Nuclear Information System (INIS)

    Within the European regional TC Project a workshop on Planning and Management Support for NPP Personnel SAT Based Training programmes was held at the Paks NPP Maintenance Training Centre in June 1998. It was organized jointly by IAEA and the Paks NPP. The objective of the workshop was to present the important experience gained by successful implementation of the IAEA Model Project on Strengthening Training for Safe Operation at Paks NPP as well as international experience in developing and carrying out the projects to upgrade NPP personnel training in particular maintenance training, and competence based Systematic Approach to Training (SAT). Twenty five participants from Armenia, Bulgaria, China, Czech republic, Hungary, Romania, Russia, Slovak Republic and Ukraine attended the workshop presenting and exchanging experiences in implementing NPP SAT-based training programmes

  13. Risk Informed Support of Decision Making in Nuclear Power Plant Emergency Zoning - Generic Framework towards Harmonising NPP Emergency Planning Practices

    OpenAIRE

    KUBANYI JOZEF; BOLADO LAVIN RICARDO; SERBANESCU DAN; TOTH BELA; WILKENING HEINZ

    2008-01-01

    The report provides a systematic overview of the essential aspects of risk informed support of decision making (RIDM) in nuclear power plant (NPP) emergency zoning (EZ) as a contribution to harmonising strategic practices in the area. Owing to the state-of-the-art understanding and increased characterisation of NPP severe accidents, overall management of them should be analysed as an integrated complex process. The interrelationship of NPP emergency operating procedures, safety and risk asses...

  14. On the foundation of new ''Hygienic guidlines on NPP design and operation'' (SP-AEhS-615/9-79)

    International Nuclear Information System (INIS)

    New ''Sanitary guidlines for NPP design and operation'' written considering requirements of RPG - 76 and Main sanitary guidlines-72 are discussed. The following chapters are characterized in detail namely, requirements for dose for personnel and population inhabitating the NPP vicinity; population doses at maximum design accident; estimated effluents corresponding to accidental radiation doses, gaseous aerosol effluents during NPP normal operation reguirements for liquid and solid radioactive waste disposal. Tabular data are presented

  15. Modeling of temporal and spatial coherency of net primary production (MODIS NPP) in the mountain forests of South Siberia

    Science.gov (United States)

    Ivanova, Yulia; Soukhovolsky, Vlad

    Net primary production (NPP) of mountain forest is very variable and depends on a variety of external modifying factors such as intensity and spectrum of solar radiation, climatic conditions in the area. Less studied are features of long-term NPP dynamics associated with self-regulation processes of tree growth in a forest. Mountain forests are a convenient object for analysis and modeling of long-term NPP changes that do not depend on climatic factors, since in mountain forests climatic conditions are uniquely determined by altitude Temporal and spatial coherence of mean annual NPP time series (Yrs. 2000 - 2012) was studied according to data from satellite observations of MODIS/TERRA. Mean annual NPP estimates' series were examined for different altitudinal zones in the Sayan Mountains (South of Central Siberia). Altitudes ranged from 600 to 1.800 meters above sea level. This area is lengthful vertically and has well-marked mountain-belt vegetation complexes, where mixed forests, fir and pine coniferous forests, alpine meadows and alpine tundra successively come one after another. Spatial and temporal coherence of NPP time series for different habitats is analyzed. The analysis showed that variations in annual NPP values of the fir (Abies sibirica) at different altitudes (450-1700 m) are synchronized. These variations in NPP values are described by the AR(2) model. Such behavior of studied NPP time series suggests a lag in growth of woody plants. In this case, the current NPP is influenced by the NPP values of the two previous years. At higher altitudes, where the density of the trees decreases and herbaceous plants become dominant, the range and synchronization of NPP variations decrease.

  16. The chemical decontamination of the Callisto PWR loop

    International Nuclear Information System (INIS)

    The CALLISTO (Capability for Light water Irradiation in Steady state and Transient Operation) is a PWR experimental facility for scientific in-pile studies installed into the BR2 Material Test Reactor. Three experimental rigs, called In-Pile Sections (IPS), are installed in three reactor channels. They are connected to a common pressurized loop, which operates with representative PWR water chemistry (typically 400 ppm boron, 3,5 ppm lithium and 30 ccSTP/kg dissolved hydrogen). The IPSs can be provided with adequate instrumentation and be modified to perform valid irradiation studies in a high neutron flux and in a relevant thermos-hydraulic environment. During more than 15 years of operation, activation products have accumulated into the loop leading to a continuous increase of the dose rates at the work area. Consequently periodic maintenance and inspection operations have become more and more expensive in terms of collective dose uptake. In consultation with the internal and external safety authorities the decision has been made to proceed to the chemical closed-loop decontamination of the most important components of CALLISTO (heater, pressurizer, main and bleed flow coolers). The objective of reducing the dose rates without compromising the integrity of the operational loop has led to the combined use of known soft chemical decontamination products as KMnO4 and H2C2O4. About 10 GBq of Co-60 activity and 250 g of corrosion products were removed from the stainless steel CALLISTO loop. The systems involved had a total volume of 0,5 m3 and a surface area of 18 m2. All released activity and corrosion products were removed by ion exchange resins, leading to the generation of 2x150 liters of radioactive waste. The dose rate reduction factors in contact with the treated components varied between 2 and 12. The collective dose uptake of the entire operation (preparation - decontamination - clean-up) was about 5,5 man.mSv, and thereby in line with the ALARA estimations

  17. Issues and remedies for secondary system of PWR/VVER

    International Nuclear Information System (INIS)

    Secondary side degradation of steam generators (SG) and Flow Accelerated Corrosion (FAC) in the secondary system have been for a long time important issues in PWR and VVER types of Nuclear Power Plants. With the evolution of the design, the most important issues are progressively moving from secondary side corrosion of Alloy 600 SG tubing, which is being replaced, to a larger variety of risks associated with potential inadequate chemistries. As far as FAC of carbon steel is concerned, the evolution of treatment selection for minimizing corrosion products transport toward the SG, as well as progressive replacement of components in the feedwater train, decreases the risk of dramatic failures which have occurred in the past. After having briefly explained the reason for the past problems encountered in the secondary system of PWR and VVER, this paper evaluates the risk associated with various impurities or contaminants that may be present in the secondary system and how to mitigate them in the most appropriate, efficient, economical and environmental friendly way. The covered species are sodium, calcium, magnesium, chloride, sulfate and sulfur compounds, fluorides, organic compounds, silica, oxygen, lead, ion exchange resins. This paper also proposes the best remedies for mitigating the new issues that may be encountered in operating plants or units under construction. These are mainly: - Selecting a steam water treatment able to minimize the quantity of corrosion products transported toward the SG; - Mitigating the risk of Flow Induced Vibration by a proper control of deposits in sensitive areas; - Minimizing the risk of concentration of impurities in local areas where they may induce corrosion; - Avoiding the presence of abnormal quantities of some species in SG, such as the detrimental presence of lead and ion exchange resin debris or the controversial presence of organic compounds; - Optimizing costs of maintenance activities (SG mechanical, chemical cleaning

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  19. Impact of climate aggregation over different scales on regional NPP modelling

    Science.gov (United States)

    Kuhnert, Matthias

    2016-04-01

    Model input data aggregation methods and data aggregation across spatial scales affect various model outputs, e.g. Net Primary Productivity (NPP). The scale at which data is collected is of great importance. In ecosystem modelling studies we often see soil and climate data collected at coarse scale being used in models to predict ecosystem responses e.g. NPP in dependency of these parameters at finer scale. Outputs of these models are impacted by the way the data is aggregated or dis-aggregated to the spatial scale. Up to know there are very few studies which quantified the impact of scaling on the simulation results. In this study, we quantify the impact of climate data aggregation using five different resolutions, to simulate NPP by 11 different crop and biogeochemical models for the same study area. The aggregation effect is investigated for wheat and maize cropping systems in the state of North Rhine-Westphalia, Germany. The simulation results are analysed for NPP averaged over growing seasons of a 30 year period at different spatial resolutions as well as for annual NPP during growing season. While there is only a minor impact of input data aggregation on NPP on 30 year averages, the annual data show differences in NPP up to 9.4 % and 13.6 % between the different resolutions for wheat and maize, respectively. The scale effect differ between the models and shows higher impacts for extreme years. This is tested by selecting years with extreme dry conditions based on a drought index, which showed stronger scale effects of up to 12.8 % and 15.5 % for wheat and maize, respectively.

  20. Managing RTP Console Upgrading Project: Best Practice for Nuclear Malaysia as TSO in Supporting NPP Development

    International Nuclear Information System (INIS)

    Human Resource Development (HRD) is required for Nuclear Power Programme (NPP). To be a Technical Support Organisation (TSO) for NPP, Nuclear Malaysia should be ready to take the responsibility in supporting Nuclear Regulatory Agency (NRA) and NPP Operators. In nurturing Nuclear Malaysia as TSO, the prime important and focus of HRD for the NPP is the reactor engineering technology. Nuclear Malaysia gives various phases of supports needed to build NPP such as during siting, design, planning, licensing, construction, commissioning, operation and maintenance in its own way and capability. The current Nuclear Malaysia unique approach is the TRIGA PUSPATI reactor (RTP) upgrading project. Research reactor plays an important role in Research and Developpement organization as a nuclear facility to assist the development of NPP. Therefore, upgrading the research reactor is needed to build the skills and gain knowledge of workers to work safely. After 29 years of operation, the RTP system is facing aging problems due to many components in the reactor are outdated. Therefore, immediate action should be carried out to mitigate the aging factor of the reactor to prevent the worsening of the aging problem, and to prevent untoward incident from happening. Action should also cover short and long term planning to prevent current situation from recurring. Currently, RTP is upgrading its console from analog to digital system. One of the achievements in this console upgrading project is the development and implementation of project management. This paper comprises the overview on the RTP console upgrading project, the project management and how this project can lead Nuclear Malaysia to be a good TSO for the development of NPP. (author)

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Indian Academy of Sciences (India)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

  5. Reassessment of PWR pressure-vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    A continuing analysis of the PTS problem associated with PWR postuated OCA's indicates that the previously accepted degree of conservatism in the fracture-mechanics model needs to be more closely evaluated, and if excessive, reducted. One feature that was believed to be conservative was the use of two-dimensional as opposed to finite-length (three-dimensional) flaws. A flaw of particular interest is one that is located in an axial weld of a plate-type vessel. For those vessels that suffer relatively high radiation damage in the welds, the length of the flaw will be no greater than the length of the weld, and recent calculations indicate that a deep flaw of that length (approx. 2 m) is not effectively infinitely long, contrary to previous thinking. The benefit to be derived from consideration of the 2-m flaw and also a semielliptical flaw with a length-to-depth ratio of 6/1 was investigated by analyzing several postulated transients. In doing so the sensitivity of the benefit to a specified maximum crack arrest toughness and to the duration of the transient was investigated. Results of the analysis indicate that for some conditions the benefit in using the 2-m flaw is substantial, but it decreases with increasing pressure, and above a certain pressure there may be no benefit, depending on the duration of the transient and the limit on crack arrest toughness

  6. Applicability of oxygenated water chemistry for PWR secondary systems

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H.P. [Studsvik Nuclear AB, Nykoeping (Sweden); Takiguchi, H.; Otoha, K. [Japan Atomic Power Co., Tokyo (Japan)

    2002-07-01

    Introduction of oxygenated water chemistry (OWC) in PWR secondary side is considered as a means to reduce the transportation of corrosion products into the steam generator and thus also minimizing crevice deposits and subsequent materials problems. One main concern, however, is the risk of inter-granular attack (IGA) in crevices. In order to study effects on crevice tube IGA by OWC, a series of experiments were performed in a steam generator (SG) simulating loop. This comprised a SG tube and a tube support plate (TSP) together forming the crevice. The over-all objective of the work accounted here was to demonstrate that it is possible to operate the steam generator secondary side with OWC without causing intolerable IGA or other types of attack on the tube in the crevice area. Tubes of sensitized Alloy 600 were exposed during a total of nine experiments in an autoclave using a TSP/tube arrangement with an asymmetric crevice design. Experiments were performed at high and low pH and potential under open and packed crevice conditions. The aggressiveness of the crevice environment was also further increased by addition of carbonate and chloride. Furthermore the tube was pressurized. Experimental parameters were monitored on the primary side as well as in the secondary bulk phase and in the crevice. (authors)

  7. Enhancing heat transfer and crud mitigation in PWR fuel

    International Nuclear Information System (INIS)

    This paper discusses three methods for increasing single phase heat transfer in PWR fuel. The primary effect of increasing heat transfer is a reduction in the steaming rate from the fuel rods, which in turn reduces the likelihood of crud formation on the fuel rods and the potential for adsorption of boron into the crud. The advantage of lowering boron mass on the fuel is reduced risk of Axial Offset Anomaly (AOA). Another benefit of reduced crud formation is a lower risk of localized corrosion, a known contributor to rod cladding failures. Thinner crud leads to locally lower rod operating temperatures (lower corrosion rate) since crud acts as a thermal insulator between the rod and the coolant. The first method of increasing heat transfer involves addition of more than one Intermediate Flow Mixing vane grid (IFM) in the span between two neighboring structural spacing grids. The second method includes optimization of the mixing vane according to axial position. The third method involves variation of the IFMs axial position to optimize axial distribution of rod heat transfer. (authors)

  8. Boron mixing transient in a PWR vessel. Physical studies

    International Nuclear Information System (INIS)

    EDF has conducted a R and D action, aiming at gaining more knowledge on vessel thermal-hydraulics; it consists of two complementary approaches based on mock-up experiments and numerical simulations. Maintenance scenarios studies began in 1995. They have been performed solely with the FEM CFD code N3S. The FEM model take into account the U pipe, the primary pump and the cold leg. This mesh can be connected to the vessel mesh used in the study of previous configurations. The first case in progress concerns the influence of the start-up of a boron unsaturated demineralizer. The study concerns the plug formation in the U pipe involved by the clear and cold seal injection water entering the primary circuit. At the end of the diluted water injection the primary pump is started up and the U pipe fluid is sent in the reactor vessel. This paper presents first the CPY 900 MW PWR vessel taken into account in these physical studies, with a special focus on the geometric peculiarities. Then the 1/5. scale BORA-BORA mock-up and the 3D FEM Thermal Hydraulic code N3S are described. The results obtained until now are presented. The degree of achievement of the studies on the three priority cases (start-up, hot shut-down normal operation, cold shut-down normal operation)

  9. PWR loading pattern optimization using Harmony Search algorithm

    International Nuclear Information System (INIS)

    Highlights: ► Numerical results reveal that the HS method is reliable. ► The great advantage of HS is significant gain in computational cost. ► On the average, the final band width of search fitness values is narrow. ► Our experiments show that the search approaches the optimal value fast. - Abstract: In this paper a core reloading technique using Harmony Search, HS, is presented in the context of finding an optimal configuration of fuel assemblies, FA, in pressurized water reactors. To implement and evaluate the proposed technique a Harmony Search along Nodal Expansion Code for 2-D geometry, HSNEC2D, is developed to obtain nearly optimal arrangement of fuel assemblies in PWR cores. This code consists of two sections including Harmony Search algorithm and Nodal Expansion modules using fourth degree flux expansion which solves two dimensional-multi group diffusion equations with one node per fuel assembly. Two optimization test problems are investigated to demonstrate the HS algorithm capability in converging to near optimal loading pattern in the fuel management field and other subjects. Results, convergence rate and reliability of the method are quite promising and show the HS algorithm performs very well and is comparable to other competitive algorithms such as Genetic Algorithm and Particle Swarm Intelligence. Furthermore, implementation of nodal expansion technique along HS causes considerable reduction of computational time to process and analysis optimization in the core fuel management problems

  10. Robots in P.W.R. nuclear powerplants

    International Nuclear Information System (INIS)

    The satisfactory operation of 37 900-MWe PWR powerplants in France, Belgium and South-Africa and the start-up of 1300 MWe powerplants allowed the development of a wide range of automatic units and robots for the periodic maintenance of nuclear plants, reducing the risk of ionizing radiation for the personnel. A large number of automated tools have been built. Among them: - inspection and maintenance systems for the tube bundle of steam generators, - robotized arms ROTETA and ROMEO for the heavy maintenance and delicate operations such as tube extraction or shot peening of tubes to improve their resistance to corrosion; - the versatile manipulator T.A.M. with electrically controlled articulations. The development of functionally versatile tools and robots and the integration of new technologies such as 3-D vision allowed the construction of the self-guided vehicle FRASTAR capable of moving within a nuclear building and in a cluttered environment. This vehicle includes means for avoiding isolated obstacles and can move on stairs

  11. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Directory of Open Access Journals (Sweden)

    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.

  12. Integrated training support system for PWR operator training simulator

    International Nuclear Information System (INIS)

    The importance of operator training using operator training simulator has been recognized intensively. Since 1986, we have been developing and providing many PWR simulators in Japan. We also have developed some training support systems connected with the simulator and the integrated training support system to improve training effect and to reduce instructor's workload. This paper describes the concept and the effect of the integrated training support system and of the following sub-systems. We have PES (Performance Enhancement System) that evaluates training performance automatically by analyzing many plant parameters and operation data. It can reduce the deviation of training performance evaluation between instructors. PEL (Parameter and Event data Logging system), that is the subset of PES, has some data-logging functions. And we also have TPES (Team Performance Enhancement System) that is used aiming to improve trainees' ability for communication between operators. Trainee can have conversation with virtual trainees that TPES plays automatically. After that, TPES automatically display some advice to be improved. RVD (Reactor coolant system Visual Display) displays the distributed hydraulic-thermal condition of the reactor coolant system in real-time graphically. It can make trainees understand the inside plant condition in more detail. These sub-systems have been used in a training center and have contributed the improvement of operator training and have gained in popularity. (author)

  13. Dynamic modelling of PWR fuel assembly for seismic behaviour

    International Nuclear Information System (INIS)

    Vibration and snap back tests have shown that the behaviour of PWR fuel assemblies was non linear : the fuel assembly eigenfrequencies decrease with the excitation level or with the motion amplitude, which was supposed to be due to the slippage of the fuel rods through the grids. Up to now the fuel assembly models were linear and composed by one beam alone representing both the guide thimbles and the fuel rods or by two beams (one for the guide thimbles and one for the fuel rods). The stiffness of such models' were adjusted to fit with the measured eigenfrequency corresponding to a given amplitude. The aim of this paper is to identify the influence of the slippage between grids and fuel rods on the dynamic behaviour of the fuel assembly. For that purpose a non linear fuel assembly model is proposed representing explicitly the slippage phenomenon and is applied to the reduced scale fuel assemblies which have been tested in the framework of a collaboration between FRAMATOME and CEA-DMT. Comparisons between calculations and experiments will be presented and the limitation of this model will be also discussed

  14. Break location effects on PWR small break LOCA phenomena

    International Nuclear Information System (INIS)

    The report presents experimental results of a small lower plenum break test of SB-PV-01 conducted at the large-Scale Test Facility (LSTF) of the Rig-of-Safety Assessment (ROSA)-IV program. This test simulates a loss-of-coolant accident (LOCA) caused by instrument tubes break (break area corresponds to 0.5% of the cold leg flow area) in a Westinghouse-type pressurized water reactor (PWR) assuming both manual actuation for all of the high pressure injection (HPI) systems and failure of the auxiliary feedwater systems. The report clarifies long-term system responses, especially the core cooling conditions related to the primary mass inventory. Also it clarifies break location effects on small break LOCA phenomena by comparing other five similar LOCA tests with break locations at cold leg, hot leg, upper head, pressurizer top (TMI-type) and SG U-tubes. It is coucluded that the lower plenum break is the severest on core heatup due to the highest break flow rate and the least primary mass recovery after the ECCS among the six tests. (author)

  15. Automatic defect identification on PWR nuclear power station fuel pellets

    International Nuclear Information System (INIS)

    This article presents a new automatic identification technique of structural failures in nuclear green fuel pellet. This technique was developed to identify failures occurred during the fabrication process. It is based on a smart image analysis technique for automatic identification of the failures on uranium oxide pellets used as fuel in PWR nuclear power stations. In order to achieve this goal, an artificial neural network (ANN) has been trained and validated from image histograms of pellets containing examples not only from normal pellets (flawless), but from defective pellets as well (with the main flaws normally found during the manufacturing process). Based on this technique, a new automatic identification system of flaws on nuclear fuel element pellets, composed by the association of image pre-processing and intelligent, will be developed and implemented on the Brazilian nuclear fuel production industry. Based on the theoretical performance of the technology proposed and presented in this article, it is believed that this new system, NuFAS (Nuclear Fuel Pellets Failures Automatic Identification Neural System) will be able to identify structural failures in nuclear fuel pellets with virtually zero error margins. After implemented, the NuFAS will add value to control quality process of the national production of the nuclear fuel.

  16. Effect of coolant chemistry on PWR radiation transport processes

    International Nuclear Information System (INIS)

    The effect of various PWR-type coolant chemistry regimes on the behavior of corrosion products has been studied in the DIDO Water Loop at Harwell. While the extent of in-core spinel deposition is influenced by pH in a manner to be expected from the temperature coefficient of solubility of nickel-iron spinel, there is evidence that boric acid plays a role apart from its influence on pH. Out-of-core deposition of active cobalt on stainless steel takes place largely in the chromium-rich inner oxide layer, and there is also significant uptake of corrosion products into the film on Zircaloy. Deposition depends on flow characteristics in different ways for different elements. The evidence suggests that in DWL soluble species are dominant in out-of-core deposition processes for corrosion products. The adsorption of cobalt in zirconium oxide provides a route for deposition on fuel elements which may in some circumstances be more significant than spinel deposition. A most important factor governing deposition behavior is surface condition; the influence of weld regions and the effect of varying pretreatment conditions have both been demonstrated. Some alternative chemistry regimes have been explored, but do not appear to offer any advantages with respect to activity transport control over the more conventional regime based on lithium hydroxide and hydrogen dosing. 8 refs., 26 figs., 28 tabs

  17. PWR-440 water chemistry optimization to reduce AOA effect

    International Nuclear Information System (INIS)

    The pressure drop increase in PWR-440 is mainly caused by the fact that the coolant contains numerous corrosion products, which are generated after decontamination and deposited in the top part of the fuel assembly as well as by coolant nucleate boiling that under standard water chemistry conditions leads to acceleration of corrosion products deposition and coolant radioactivity growth respectively. The modeling of the pressure drop changes were based on standard data of water chemistry, reactor operating characteristics and fundamental thermodynamic parameters to predict the pressure drop growth. The results of the performed research and modeling of the corrosion products mass transfer processes allowed to qualify relative contribution of thermohydraulic and chemical parameters in the processes and to fulfill the activities as follows: To perform power units operation at water chemistry with maximum permissible alkali metals content. To increase the coolant flow rate through the core; to do so, throttling orifices were replaced and canister-shields were removed. To reduce the number of steam generators to be decontaminated to 2 per year in a single power unit. As a result deposits accumulation in fuel assemblies has been minimized and there is no leakage in the fuel element; reactor thermal output limitation has been eliminated. (author)

  18. Development of a dry storage cask for PWR spent fuel

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power Co., Ltd.(KHNP), which operates all the nuclear power plants in Korea, is developing a new dry storage cask to store twenty four spent fuel assemblies generated from pressurized water reactors for at-reactor or away-from-reactor interim storage facility in Korea. The dry storage cask is designed and evaluated according to the requirements of the IAEA, the US NRC and the Korean regulations for the dry spent fuel storage system. It provides confinement, radiation shielding, structural integrity, subcritical control and passive heat removal for normal and accident conditions. The dry storage cask consists of a dual purpose canister providing a confinement boundary for the PWR spent fuel, and a storage overpack providing a structural and radiological boundary for long-term storage of the canister placed inside it. The overpack is constructed by a combination of steel and concrete, and is equipped with penetrating ducts near its lower and upper extremities to permit natural circulation of air to provide for the passive cooling of the canister and the contained spent fuel assemblies. This paper describes development status, description, design criteria, evaluation and demonstration tests of the dry storage cask. (authors)

  19. Barium silicate glass/Inconel X-750 interaction. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kelsey, Jr., P. V.; Siegel, W. T.; Miley, D. V.

    1980-01-01

    Water reactor safety programs at the Idaho National Engineering Laboratory have required the development of specialized instrumentation. An example is the electrical conductivity-sensitive liquid level transducer developed for use in pressurized-water reactors (PWRs) in which the operation of the sensing probe relies upon the passage of current through the water between the center pin of the electrode and its shell such that when water is present the resulting voltage is low, and conversely, when water is absent the voltage is high. The transducer's ceramic seal is a hot-pressed glass ceramic; its metal housing is Inconel X-750. The ceramic material provides an essential dielectric barrier between the center pin and the outer housing. The operation of the probe as well as the integrity of the PWR environment requires a hermetically-bonded seal between the ceramic and the metal. However, during testing, an increasing number of probe assemblies failed owing to poor glass-to-metal seals as well as void formation within the ceramic. Therefore, a program was initiated to characterize the metallic surface with respect to pre-oxidation treatment and determine optimum conditions for wetting and bonding of the metal by the glass to obtain baseline data relevant to production of acceptable transducer seals.

  20. PWR safety/relief valve blowdown analysis experience

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Advanced methods for the study of PWR cores

    International Nuclear Information System (INIS)

    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.)

  2. Knowledge management at Kozloduy NPP - challenges and solutions

    International Nuclear Information System (INIS)

    Like any highly technical endeavor, the use of nuclear technology relies heavily on a vast accumulation of knowledge - volume upon volume of scientific research, engineering analysis, design documentation, operational data, maintenance records, regulatory reviews and countless other pieces of technical information, combined with complex reservoir of scientists, engineers, technicians and managers with relevant background and expertise. The effective management of nuclear knowledge includes systematic and organized application of accumulated specific knowledge as well as ensuring the sustainable availability of qualified personnel. This means that: 1) The know-how should be properly identified, captured and classified; 2) Adequate tools for support, access and transfer of data should be available; 3) Renewing of workforce without loss of institutional memory should be guaranteed. During the centuries this process has been applied spontaneously and there are a lot of examples that the world use to be workable earlier too. However nowadays this type of spontaneous knowledge management is ineffective. Information volume and the communication efficiency have increased so much that new tools have to be created to handle the knowledge being communicated trough numerous sources. Besides information space is strongly polluted. The information channels provide unnecessary redundant information. Target oriented searching and delivering of useful information is difficult and requires significant resources. What value this know-how represent, how can it be maintained for and passed to the next generation, what are the problems of aging and renewing of workforce, what measures have been taken by Kozloduy NPP and what are the challenges to be faced yet - these are the issues to be addressed in this paper. (authors)

  3. Radioecological evaluations of Ignalina NPP Lithuania Normal operating

    International Nuclear Information System (INIS)

    The Ignalina NPP (INPP) with two RBMK-1500 reactors is constructed in north-eastern part of Lithuania, close to its borders with Latvia and Belarus. The natural Lake Drukshiai serves as a cooling basin for INPP. Environmental monitoring in INPP region is conducted since 1979. It was extended since 1993 due to implementation of State Scientific Research Programme INPP and Environment. Radionuclides transport in both aquatic and terrestrial ecosystems in this time was studied especially comprehensively. Accumulation of artificial radionuclides in lake is governed by biogeochemical transport of radionuclides from three sources, i.e. of global ones, emitted after Chernobyl accident and released from INPP. Radioactive releases from INPP into lake formed small activity fraction of different radionuclides in water, bottom sediments as well as in biota. Radionuclides activity in water varied from 1.4 to 7 for 137 Cs, from 0.5 to 1.7 for 60 Co, from 25 to 60 Bq/m 3 for 90 Sr, and were not detected for 134 Cs and 54 Mn. Considering terrestrial ecosystem areal distribution of radionuclides in soil and biota were evaluated. Artificial radionuclides from three mentioned above sources are distributed in ecosystem which can be treated as little contaminated (mean 137 Cs activity reaches 1000 Bq/m 2 ). Local anomalies close to INPP are related to releases from water communications of INPP. Using site specific data nuclide specific dose factors for both atmospheric and aquatic pathways and individual effective doses to members of critical groups were calculated. Doses caused by radioactive releases into water and into air during 1997 are equal 0.01 mSv/a and 0.001 mSv/a, respectively

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

    International Nuclear Information System (INIS)

    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

  5. A GeoServices Infrastructure for Near-Real-Time Access to Suomi NPP Satellite Data

    Science.gov (United States)

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

    2012-12-01

    The new Suomi National Polar-orbiting Partnership (NPP) satellite extends NASA's moderate-resolution, multispectral observations with a suite of powerful imagers and sounders to support a broad array of research and applications. However, NPP data products consist of a complex set of data and metadata files in highly specialized formats; which NPP's operational ground segment delivers to users only with several hours' delay. This severely limits their use in critical applications such as weather forecasting, emergency / disaster response, search and rescue, and other activities that require near-real-time access to satellite observations. Alternative approaches, based on distributed Direct Broadcast facilities, can reduce the delay in NPP data delivery from hours to minutes, and can make products more directly usable by practitioners in the field. To assess and fulfill this potential, we are developing a suite of software that couples Direct Broadcast data feeds with a streamlined, scalable processing chain and geospatial Web services, so as to permit many more time-sensitive applications to use NPP data. The resulting geoservices infrastructure links a variety of end-user tools and applications to NPP data from different sources, and to other rapidly-changing geospatial data. By using well-known, standard software interfaces (such as OGC Web Services or OPeNDAP), this infrastructure serves a variety of end-user analysis and visualization tools, giving them access into datasets of arbitrary size and resolution and allowing them to request and receive tailored products on demand. The standards-based approach may also streamline data sharing among independent satellite receiving facilities, thus helping them to interoperate in providing frequent, composite views of continent-scale or global regions. To enable others to build similar or derived systems, the service components we are developing (based in part on the Community Satellite Processing Package (CSPP) from

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  7. Effect of ethanolamine injection on wall thinning rate of PWR carbon steel components

    International Nuclear Information System (INIS)

    For pH control of PWR secondary system water chemistry, some plants have changed to ethanolamine injection. The purpose of this work was to understand the effect of changing water chemistry on wall thinning rate of the PWR secondary system due to flow accelerated corrosion. For that purpose, evaluations of water chemistry were carried out by a mass balance calculation, wall thinning rate measurement by a rotary disk test and wall thinning rate evaluation based on calculated magnetite solubility. As a result, it was found to be effective to inhibit the wall thinning rate of the PWR secondary system due to flow accelerated corrosion by ethanolamine injection, but it was not sufficiently effect to neglect wall thinning rate due to flow accelerated corrosion. The effect of the wall thinning rate inhibition also varied greatly for each component of the PWR secondary system. It was found that maintenance of the carbon steel used for the PWR secondary system was still required under ethanolamine injection condition. (author)

  8. Analysis on the effect of local content to the economic aspect of NPP

    International Nuclear Information System (INIS)

    Study analysis on the effect of local content to the economic aspect of NPP has already been conducted. This scope of study covers: NPP technology system analysis (OPR-1000), national industry capability to participate in NPP construction and effect analysis toward the economic. Economic analysis is conducted for 2 units NPP OPR-1000. This study concludes that: (1) The level of local content for the first 2 units NPP in Indonesia about 35.34% (pessimistic case) and 43.38% (optimistic case). (2) To reach the local content government support is needed, i.e. policy of industry, particularly the police related to technology transfer. (3) The total generation cost that could be saved per year in pessimistic case is around US$ 14.55 million - US$ 14.57 million and in optimistic case it is around US$ 22.23 million – US$ 22.26 million. (4) The local content implementation will provide decreasing of generation cost about 1.75% up to 1.76% in pessimistic case, and around 2.67% up to 2.68% in optimistic case. (author)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Studies of the landscape and its components of the Ignalina NPP region

    International Nuclear Information System (INIS)

    The causes of landscape degradation in Lake Drukshiai basin are the following: building and operation of Ignalina NPP and Visaginas town with the service infrastructure. Landscape degradation map was compiled showing the areas effected by degradation, intensity of degradation and resistance of landscape against this process. It was determined that the highly de gradated landscape in the lake basin covers the area of 395.3 ha (1.43% without the lake). The landscape resistance to the processes of degradation depends on the migration fluxes of erosion products. The investigation revealed that erosion products in Lake Drukshiai basin are the least stable in the landscape with fluvioglacial sand and gravel deposits. The land flora in Ignalina NPP region is being represented by 617 plant species belonging to 81 family. Temperature, chemical and physical changes which occurred in the lake - cooler after the Ignalina NPP was put in operation were primarily evident in the structure and abundance of by-shore complexes of micro arthropods and insects.During the first five years of operation of Ignalina NPP the abundance and diversity of species at the shores of the lake increased 2.5 - 3.5 times. The genotoxic al impact of Ignalina NPP environment on local animal populations during 184 - 1996 period was slight and did not change significantly. (author)

  12. Analysis of an NPP Structure subjected to Vibrations Induced from Airplane Crashes

    International Nuclear Information System (INIS)

    After the terrorists' attacks with civilian airplanes on September 11, 2001, special attention has been paid to the potential for an airplane crash into a Nuclear Power Plant (NPP) as a man-made hazard. An airplane crash (APC) into an NPP has the potential to damage the roofs and walls of these structures, as well as other systems and components such as pipelines, electric motors, power supplies, power cables of electricity transmission that are important for safety. Therefore, an evaluation of the structural response to an APC is important for the safety of NPPs to be confirmed. A structural integrity analysis was carried out focusing on the vibration effects of an APC on an NPP structure. The NPP structure under consideration has been conceptually redesigned based on APR1400 to have double containments for the purpose of a feasibility study to meet European requirements. The finite element method was used for the structural analysis of the NPP, and the computer code ABAQUS was employed for this analysis

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

    Directory of Open Access Journals (Sweden)

    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

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

    Science.gov (United States)

    Campioli, M.; Gielen, B.; Göckede, M.; Papale, D.; Bouriaud, O.; Granier, A.

    2011-09-01

    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 contribution of previous-year reserves to stem growth, as well as reduction

  15. Survey of experiments and code development for the passive residual heat removal system of PWR in China

    Institute of Scientific and Technical Information of China (English)

    HUANG Yan-Ping; ZHUO Wen-Bing; YANG Zu-Mao; XIAO Ze-Jun; CHEN Bing-De; JIA Dou-Nan

    2004-01-01

    Three different kinds of experiments and their typical results are surveyed for the passive residual heat removal system (PRHRS) of PWR performed in Nuclear Power Institute of China (NPIC) recent ten years. The typical results of MISAP. a special code for PWR passive residual heat removal system developed and assessed by NPIC,are also described briefly in this paper.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Institute of Scientific and Technical Information of China (English)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Probabilistic assessment of the Juragua NPP response under Station Blackout conditions

    International Nuclear Information System (INIS)

    Assessment of the NPP response under SBO (station Blackout) conditions is a current safety issue of special interest, In the case of Juragua NPP, the safety assessment related to this topic is very important, taking into account the peculiarities of the Cuban Electro energetic System: small and long island, without possibilities of conexion beyond its borders and under the incidence of tropical phenomena In this papers a preliminary evaluation is presented of the potential incidence of Station Blackout conditions for Juragua NPP. the importance of this sort of events for the safety of the plant is evaluated, the factors which condition it are identified and measures for its prevention or recovering the normal situation if such an event takes place are proposed

  20. Principles of arranging a specialized operating system for the NPP APCS

    International Nuclear Information System (INIS)

    This paper discusses problems related to the development of software for the low level computers of nuclear power plant (NPP) automated process control systems (APCS). These systems monitor and control process equipment and are connected to the NPP control panel. The authors propose concepts and principles for developing specialized system software. The NPP APCS has a two to three level structure. For the upper level it is appropriate to use more generalized software (operating systems, network support, databases, etc.), but for lower levels it is necessary to develop specialized software to meet often conflicting requirements of hardware configuration and functions. Details of developing an operating system for a OVER-500 power unit are presented. (L.L.)