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Sample records for 300mwe pwr npp

  1. Qinshan 300Mwe NPP full scope simulator upgrade

    On April 28,2004, RINPO was awarded the project for Qinshan 300Mwe NPP full scope simulator upgrade, the SAT (site acceptance test) was completed on June 30 2005 and the simulator put into operator training again. Scope of upgrade includes: computer system (DGI server and workstations) all replaced by microcomputers; G2 I/O controllers all replaced by RTP EIOBC; Unix-based simulation support environment replaced by RINPO's PC-based simulation environment RINSIMTM, Instructor software replaced by RINPO's PC-based instructor software with function and diagram redesigned; DEH, Feed-water control and some other digital control systems redeveloped to follow NPP modifications; desk-top simulator with soft panel control room developed as byproduct; most of the models not changed but it is planned the reactor core and PPC model will be upgraded in near future. SAT of upgrade demonstrates that the performance of the simulator much improved after the upgrade. (author)

  2. Study and economics analysis for 18-month refueling management on power uprate of a 300 MWe NPP

    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 300 MWe NPP after power uprate. The results show that the concluded fuel management scheme for a 300 MWe NPP after power uprate achieves the projected 18- month refueling cycle design objectives with the nominal thermal power of 1250 MW and meets the design criteria. As compared to the current fuel management strategy of a 300 MWe 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 300 MWe NPP and optimization of fuel management strategy. (authors)

  3. Structural mechanics research and development for main components of Chinese 300 MWe PWR NPPs: from design to life management

    Qinshan Nuclear Power Plant (Unit I), is a 300 MWe prototype PWR independently developed by Chinese own efforts, from design, manufacture, construction, installation, commissioning, to operation, inspection, maintenance, ageing management and lifetime assessment. Shanghai Nuclear Engineering Research and Design Institute (SNERDI) has taken up with and involved in deeply the R and D to tackle problems of this type of reactor since very beginning in early 1970s. Structural mechanics is one of the important aspects to ensure the safety and reliability for NPP components. This paper makes a summary on role of structural mechanics for component safety and reliability assessment in different stages of design, commissioning, operation, as well as lifetime assessment on this type PWR NPPs, including Qinshan-I and Chashma-I, a sister plant in Pakistan designed by SNERDI. The main contents of the paper cover design by analysis for key components of NSSS; mechanical problems relating to safety analysis; special problems relating to pressure retaining components, such as fracture mechanics, sealing analysis and its test verifications, etc.; experimental research on flow-induced vibration; seismic qualification for components; component failure diagnosis and root cause analysis; vibration qualification and diagnosis technique; component online monitoring technique; development of defect assessment; methodology of aging management and lifetime assessment for key components of NPPs, etc. (authors)

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

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

  5. 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

    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

  6. Comprehensive research on sealing behaviour of reactor vessel of 300 MWe nuclear power plant

    The general conception of a special research on sealing behaviour of PWR vessel is described and the major results centering on the establishment of sealing analysis program system and its experimental verification, along with the description on the development and measurement of sealing ring, the thermal sealing test and the relevant analysis are given. On the basis of the above approach, the vessel sealing behaviours of 300 MWe Qinshan Nuclear Power Plant are evaluated. A concept on the classification of pressure vessels and their sealing criteria are proposed. Two viewpoints on the analysis are suggested, which are that the vessel sealing deformation analysis should be regarded as a basis of the general stress analysis and that bolt loading increment caused by the bolt temperature lag should be taken as a key point when considering the thermo-contact coupling in transient sealing analysis. The understanding about the sealing mechanism are expounded and the thermal equivalent of hydrostatic test is discussed

  7. Affecting factors analysis of major equipment erection key path in PWR NPP

    The affecting factors of major equipment erection in PWR NPP exist impersonally, especially the design and equipment supply has produced some effects on major equipment erection of PWR nuclear power plant. Through the analysis of key path and affecting factors on major equipment erection of PWR NPP, the paper puts forward some countermeasures. (authors)

  8. Reactor building seismic analysis of a PWR type - NPP

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  9. System comparative analysis of the most advanced pressured water reactors (PWR, WWER) and boiling water reactors (BWR) projects with the aim to choose the reactors for NPP construction in Kazakhstan

    Full text: The official decision on construction of a Nuclear Power Plant (NPP) in Kazakhstan has been accepted by the Kazakhstan government. The results on the choice of the power reactors projects of the NPP are given in the report. The choice has been carried out with the aim to develop recommendation on reactors of the NPP for construction in Kazakhstan. The choice of the reactors was based on the system comparative analysis of the most advanced power reactors projects using 15 criteria system of the nuclear, radiating and ecological safety and economic competitiveness. Following Pressurized Water Reactor (PWR, WWR) projects have been subjected to the system comparative analysis: 1) Large Sized Reactors (700 MW(el) and up): such as EPR, developed by Germany Siemens and France Framatome companies; CANDU-9, heavy-water reactor, developed by Atomic Energy of Canada Ltd (AECL); System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation, developed by Korea Power Engineering Company, Inc.; APWR, Japanese advanced reactor, developed by Japan Atomic Power Company, Japan, Mitsubishi Heavy Industries, Japan and Westinghouse Electric Company, USA; WWER-1000 (V-392) - development by Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor, development by Westinghouse, USA/Genesi, Italy. 2) Medium Sized Reactors (300 MWe - 700 MWe): AP-600, passive PWR, developed by the Westinghouse company; CANDU-6, heavy-water reactor, developed by Atomic Energy of Canada Ltd (AECL); An-tilde-600, passive PWR, developed by Nuclear Power Institute of China; WWER-640, Russian passive reactor, developed by 0KB ''Gidropress'' Experimental and Design Office, Russian Federation; MS-600, developed by Mitsubishi Company; KSNP-600, developed by Korea Power Engineering Company, Inc., South Korea. 3) Small Sized Reactors (a few MWe- 300 MWe): IRIS, reactor of IV generation, developed by the International Corporation of 13

  10. Study on normalizing real-time monitoring critical safety parameters of pressurized water reactor (PWR) nuclear power plants (NPP)

    Real-time monitoring safety parameters are very important both in NPP operation surveillance and emergency management. Base on the analysis of regulations, emergency technical requirements and technical document of NPP obtained construction license before Jan 1st of 2009, the real-time Monitoring Safety Parameters of PWR NPPs in concerning range were sorted out. (authors)

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

    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.

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

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

    2010-05-15

    A new technique combining high boiler efficiency and low-NO(x) emissions was employed in a 300MWe down-fired boiler as an economical means to reduce NO(x) emissions in down-fired boilers burning low-volatile coals. Experiments were conducted on this boiler after the retrofit with measurements taken of gas temperature distributions along the primary air and coal mixture flows and in the furnace, furnace temperatures along the main axis and gas concentrations such as O(2), CO and NO(x) in the near-wall region. Data were compared with those obtained before the retrofit and verified that by applying the combined technique, gas temperature distributions in the furnace become more reasonable. Peak temperatures were lowered from the upper furnace to the lower furnace and flame stability was improved. Despite burning low-volatile coals, NO(x) emissions can be lowered by as much as 50% without increasing the levels of unburnt carbon in fly ash and reducing boiler thermal efficiency. PMID:20429548

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

    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

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

    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

  15. Seismic analysis for safety related structures of 900MWe PWR NPP

    Nuclear Power Plant aseismic design becomes more and more important in China due to the fact that China is a country where earthquakes occur frequently and most of plants arc unavoidably located in seismic regions. Therefore, Chinese nuclear safety authority and organizations have worked out a series of regulations and codes related to NPP anti-seismic design taking account of local conditions. The author presents here an example of structural anti-seismic design of 90GM We PWR NPP which is comprised of: ground motion input, including the principles for ground motion determination and time history generation; soil and upper-structure modelling, presenting modeling procedures and typical models of safety related buildings such as Reactor Building, Nuclear Auxiliary Building and Fuel Building; soil-structure interaction analysis; and in-structure response analysis and floor response spectrum generation. With this example, the author intends to give an overview of Chinese practice in NPP structure anti-seismic design such as the main procedures to be followed and the codes and regulations to be respected. (author)

  16. Design of Evaporator for Liquid Radioactive Waste Treatment-NPP 1000 MW, PWR

    The evaporator for liquid radioactive waste treatment of 1000 MW NPP-PWR has been designed. The basic calculate of this design was capacity 7000 l/hr, which 5 mg/l solid content. The system used was superheated steam 3.4 atmosphere, 281°F. The data required from design of evaporator are evaporator part (heat exchanger): diameter of shell is 33 inch (82.50 cm), nominal diameter tube is 1.5 inch (3.75 cm), number of tube is 215, tube arrangement triangular pitch is 1 inch pitch, height 600 cm; Mist separator: diameter is 200 inch (500 cm), height 600 inch (1500 cm); Condenser: diameter of shell is 33 inch (82.50 cm), nominal diameter tube is 1.5 inch (3.75 cm), number of tube is 215, tube arrangement triangular pitch is 1 inch pitch, height 600 cm. (author)

  17. Sophysticated systems for analysing standard signals of a PWR NPP for diagnostic purposes

    An expert system is presented, which was designed for WWER type nuclear power plants (NPP) with 440 MWe PWR units. The input of the expert system includes the most important technological parameters of the core and of the primary and secondary loops. The expert system consists of the reactor noise diagnostics system (RNDS) and the on-line analysis system (VERONA). RNDS processes the AC components of measured signals. The application of RNDS advanced results in the following fields: registration of base line spectra; identification and localization of in core vibration, core barrel motion, propagating disturbances and the beginning of boiling; estimation of rector parameters; sensor diagnostics. VERONA processes the DC components. The following estimates are displayed: the total power production, the power generation in each fuel assembly and at ten elevations, the heat balance. (author)

  18. Performing an 8% power uprating of Tihange 1, NPP a 900MW.PWR

    Tihange 1 NPP, an 18 year old three loop 900 MWe PWR in Belgium, is co-owned and operated by French and Belgium electric utilities. Steam Generators replacement is planned for 1995. The opportunity to simultaneously incorporate a power uprating program was investigated. Initial conclusions were : 1. The balance of plant was sufficiently overdesigned from origin to allow such an uprating; 2. Recalculation of safety margins using modern techniques released margins allowing power up rating; 3. Safety injection and auxiliary S.G. feedwater systems had to be improved; 4. The payback period for the uprating would be less than 2 years. Therefore, an uprating of 8% was programmed and new steam generators with a significant (> 25 %) increase in heat transfer area were ordered. Thermohydraulics core calculations were redone using WRBI CHF correlations and the RTDP statistical approach to redetermine DNBR. LOCA calculations are being performed with Westinghouse's new code COBRA TRACK. Neutron flux calculations particularly in determining peaking factors will probably necessitate the use of three dimensional core codes due to 15 x 15 fuel high linear power. Hardware modifications are to be carried out on the following systems: 1. Auxiliary steam generator feedwater system : replacement of the existing flow limiting orifii by venturi nozzles in the feedwater lines. 2. Safety injection system for long term LOCA: recalibration of flowrates. Final detailed engineering studies as well as hardware modifications are to be completed by mid 95. (author)

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

    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)

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

    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

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

    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

  2. Life evaluation of cast duplex stainless steel elbows in French PWR NPP

    The principal primary circuit cast elbows of French PWR are in austenitic-ferritic cast stainless steel CF8 - CF8M types. This material is sensitive to thermal aging at PWR operating temperatures. The aging results in a diminishment of tearing resistance characteristics, and with the possible presence of foundry flaws this could lead to a fear of increased break risk. An extensive program on material properties, inspection, tests in laboratory, flaw evaluations, etc, has been covered out in the last 5 years between EDF and FRAMATOME. This paper presents the major tasks performed to justify a good behaviour of these elbows, and they will remain operational at least for the 40-year design lifetime, the consequences at the maintenance level and the utility point of view. CF8 and CF8M are cast materials, that can have casting defects that we generally assume conservatively as a perfect crack for fracture mechanics analysis. The other fact is that materials can be sensitive to thermal aging that were not clearly quantified at the design level by any international code in the 70's. This paper shows EDF's maintenance strategy for those nuclear power plants at present being operated. One important task described in this paper is the material toughness evaluation work proposed to cover all the pipe elbows in 3 loops and 4 loops nuclear power plants. Presently, all the EDF PWR elbows can be used safely for the 40 year design period, some complementary work is in progress to support this conclusion. (author)

  3. A comprehensive study of PWR's primary coolant purification cures: modelling and application to a French NPP cold shutdown

    Until now, control of the effectiveness of PWR N-PP reactor primary coolant purification during cold shutdowns has mostly been provided by monitoring the purification slope of the cobalt radioisotopes, the level and activity variation of the other released radionuclides and determining the purification factors. This paper proposes a new modelling of the purification curves, more complete than those currently in force. The latter is based on the coupling of a source term (the solid in the course of dissolution) and a well term (the purification system). The simulations highlight the influence of several factors on the shape of the purification curves, such as: - the rate of release of the element in solution, - the concentration of the element in the primary coolant at the oxygenation peak, - its concentration at equilibrium with its solid state form, - the purification factor of the element, - the volume flow rate in the purification station of the Chemical and Volume Control System. The benefit of this model is illustrated by application to the results of an original measurement campaign, performed recently during the cold shutdown of Fessenheim Unit 1,900 MWe PWR NPP. A full and coherent set of measurements was taken on the primary coolant, from the Residual Heat Removal System hold point to the shutdown of the reactor coolant pumps: operating parameters, lithium concentration, pH and redox potential at 25 deg C, chemical concentrations and activities of the corrosion products determined for filtered and non-filtered samples. Theoretical primary coolant speciation has made it possible to identify the chemical forms of all the elements of interest, before and after oxygenation, and to highlight, for some of them, the likely presence of small-sized particle compounds (< 0.45 μm). Coupled with the speciation calculations, this modelling enables an interpretation to be advanced for the purification curve shapes observed and hypotheses to be proposed on the nature of

  4. Evaluation of 14C Behavior Characteristic in Reactor Coolant from Korean PWR NPP's

    This study has been focused on determining the chemical composition of 14C - in terms of both organic and inorganic 14C contents - in reactor coolant from 3 different PWR's reactor type. The purpose was to evaluate the characteristic of 14C that can serve as a basis for reliable estimation of the environmental release at domestic PWR sites. 14C is the most important nuclide in the inventory, since it contributes one of the main dose contributors in future release scenarios. The reason for this is its high mobility in the environment, biological availability and long half-life (5730 yr). More recent studies - where a more detailed investigation of organic 14C species believed to be formed in the coolant under reducing conditions have been made - show that the organic compounds not only are limited to hydrocarbons and CO. Possible organic compounds formed including formaldehyde, formic acid and acetic acid, etc. Under oxidizing conditions shows the oxidized carbon forms, possibly mainly carbon dioxide and bicarbonate forms. Measurements of organic and inorganic 14C in various water systems were also performed. The 14C inventory in the reactor water was found to be 3.1 GBq/kg in PWR of which less than 10% was in inorganic form. Generally, the activity in the water was divided equally between the gas- and water- phase. Even though organic compound shows that dominant species during the reactor operation, But during the releasing of 14C from the plant stack, chemical forms of 14C shows the different composition due to the operation conditions such as temperature, pH, volume control tank venting and shut down chemistry.

  5. The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

    Highlights: • We put forward a new PAMS and analyze its operation characteristics under SBO. • We conduct comparative analysis between PAMS and Traditional Secondary Side PHRS. • The PAMS could cope with SBO accident and maintain the plant in safe conditions. • PAMS could decrease heat removal capacity of PHRS. • PAMS has advantage in reducing cooling rate and PCCT temperature rising amplitude. - Abstract: To enhance inherent safety features of nuclear power plant, the advanced pressurized water reactors implement a series of passive safety systems. This paper puts forward and designs a new Passive Accident Mitigation Scheme (PAMS) to remove residual heat, which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Heat Removal System (PHRS). This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and analyzes the operation characteristics of PAMS to cope with the Station Blackout Accident (SBO) by using RELAP5 code. Moreover, the comparative analysis is also conducted between PAMS and Traditional Secondary Circuit PHRS to derive the advantages of PAMS. The results show that the designed scheme can remove core residual heat significantly and maintain the plant in safe conditions; the first part of PAMS would stop after 120 min and the second part has to come into use simultaneously; the low pressurizer (PZR) pressure signal would be generated 109 min later caused by coolant volume shrinkage, which would actuate the Passive Safety Injection System (PSIS) to recovery the water level of pressurizer; the flow instability phenomenon would occur and last 21 min after the PHRS start-up; according to the comparative analysis, the coolant average temperature gradient and the Passive Condensate Cooling Tank (PCCT) water temperature rising amplitude of PAMS are lower than those of Traditional Secondary Circuit PHRS

  6. Safety analysis of NPP

    This paper presents a short review of the parallel safety analysis of the various types of NPP. The NPP with PWR, WWER, BWR and HWR type reactors are mentioned. Technical, economic, location and ecology aspects of the safety of the NPP have been analysed. (author)

  7. The possibility of building nuclear power plant free from severe accident risk: PWR NPP with advanced all passive safety cooling systems (AAP SCS)

    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. (authors)

  8. Moving the largest capacity PWR dual-purpose cask in the world from Goesgen NPP to the Zwilag interim storage site

    The Swiss Goesgen nuclear power plant (NPP) has decided to use two different methods for the disposal of its spent fuel. (1) To reprocess some of its spent fuel in dedicated facilities. Some of the vitrified waste from the reprocessing will be shipped back to Switzerland using the new COGEMA Logistics, TN81 cask. (2) To ship the other part of its spent fuel to the central interim storage facility of Zwilag (Switzerland) using a COGEMA Logistics dual-purpose TN24G cask. The TN24G is the heaviest and largest dual-purpose cask manufactured so far by COGEMA Logistics in Europe. It is intended for the transport and storage of 37 pressurised water-reactor (PWR) spent fuel assemblies. Four casks were delivered by COGEMA Logistics to Goesgen NPP. Three transports of loaded TN24G casks between Goesgen and Zwilag were successfully performed at the beginning of 2002 with the new COGEMA Logistics Q76 wagon specifically designed to transport heavy casks. This article describes the procedure of operations and shipments for the first TN24G casks up to storage at Zwilag. The fourth transport of loaded TN24G was due to happen in October 2002. The TN24G cask, as part of the TN24 casks family, proved to be a very efficient solution for the KKG spent fuel management. (author)

  9. Application of the combined cycle LWR-gas turbine to PWR for NPP life extension, safety upgrade and improving economy

    Full text: The unconventional technology to extend the lifetime for the NPPs now in operation and make a construction of new NPPs cheaper - erection of steam-gas toppings to the nuclear power units - is considered in the paper. Application of the steam-gas toppings permits through reducing power of ageing reactors to extend lifetime of nuclear power unit, enhance its safety and at the same time to keep full load operation of NPP turbine and other balance-of-plant equipment. Proposed technology is examined for Russian VVER-440 reactor as an example and, also, as a pilot project, for Russian boiling VK-50 reactor now in operation Application of the steam-gas topping permits: extend the service life of ageing VVER-440 reactor by 10...15 years; use the turbine and other NPP balance-of-plant equipment at full power; increase the efficiency of combined cycle up to 48% and more; enhance the safety of NPP operation; utilize NPP balance-of-plant equipment after reactor decommissioning; perform the cost-effective operation in maneuvering modes; increase capacity factor of the plant. The construction of pilot project on the basis of the VK-50 reactor will allow not only to demonstrate new technology but also to attain appreciable economic effect including that obtained due to using the available reserves of the NPP turbine. (author)

  10. Application of the combined cycle LWR-gas turbine to PWR for NPP life extension, safety upgrade and improving economy

    The unconventional technology to extend the lifetime for the NPPs now in operation and make a construction of new NPPs cheaper - erection of steam-gas toppings to the nuclear power units - is considered in the paper. Application of the steam-ga toppings permits through reducing power of aging reactor to extend lifetime of nuclear power unit, enhance its safety and the same time to keep full load operation of NPP turbine and other balance-of-plant equipment. Proposed technology is examined for Russian VVER-440 reactor as an example and, also, as a pilot project, for Russian boiling VK-50 reactor now in operation. Heat flow sheets of the power plants, their parameters and economic problems are discussed. (author)

  11. Application of the combined cycle LWR-gas turbine to PWR for NPP life extension, safety upgrade and improving economy

    The unconventional technology to extend the lifetime for the NPPs now in operation and make a construction of new NPPs cheaper three-quarter erection of steam-gas toppings to the nuclear power units three-quarter is considered in the paper. Application of the steam-gas toppings permits through reducing power of aging reactors to extend lifetime of nuclear power unit, enhance its safety and at the same time to keep full load operation of NPP turbine and other balance-of-plant equipment. Proposed technology is examined for Russian VVER-440 reactor as an example and, also, as a pilot project. for Russian boiling VK-50 reactor now in operation. Heat flow sheets of the power plants, their parameters and economic problems are discussed. (author)

  12. Cobalt and organics removal effect using fiber filter/reverse osmosis combination process for LLRW from korean PWR NPP

    Evaporation system for liquid radioactive waste process has been used in Korean PWR nuclear power plants. The system is the most desirable process for decontamination factor (DF) theoretically. However, during the operation of the system, various problems have been arising such as scaling, carry over, etc. Because these problems make DF low, advanced technologies for liquid radwaste process have been world widely developed instead of keeping evaporation system. The main goal of new technologies is ALARA, ease of operation, cost effectiveness and minimization of environmental effect. Korea Electric Power Corporation is currently developing a combined treatment process for liquid radwaste using Micro-filter, Ultra-filter, Reverse Osmosis (RO) membrane, etc for the purpose of partly enhancement of evaporator and of having an alternative liquid radwaste process system for new reactors. As a part of the above project, the feasibility study using the Rolled Fiber-Filter (RFF) and RO membrane has been carried out. This paper reports the results of lab-test from the combined process of the fiber filtration and RO membrane module for cobalt and organics removal. The study was especially focused on the boric acid permeation in the RO unit. Because boric acid occupies large volume of the final waste after evaporation process, the new technology such as RO process has to be studied on the boron process. (author)

  13. Design safety improvements of Kozloduy NPP to meet the modern safety requirements towards the old generation PWR

    Activities related to safety improvement of Kozloduy NPP units, started at the end of 1970s included seismic resistance upgrading, fire safety improvement, reliable heat final absorber etc. During the last 10 years the approach was systematized and improved. Units 1 to 4 are of great interest; therefore here we will discuss these units only. As a result of studies and analyses performed at the end of the 1980s and the beginning of the 1990s, problems related to the safety were identified and complex of technical measures was developed and planned. A considerable part of these measures has already been implemented, and the rest will be performed during the next years. Activities were performed by stages, and at the moment the last stage is under way. It shall be finished by the year 2003. The number of the measures is quite large to describe them here in full scope -- during the first stage of the safety program (1991-1993) were developed and analyzed more than 4200 documents and more than 160 measures were executed. During the second and third stages more than 300 important improvements were realized. In the frame of the program, financed by EBRD, 10 new systems with great importance were implemented and 8 systems were significantly modified. The main measures are described below. (author)

  14. Appendix II: Adding PBMRs to a PWR site (ESKOM) - South Africa (Case study of human resource issues faced by NPP operating organizations, and how they were (or are being) addressed)

    ESKOM Holding Limited, the South African Government owned utility, operates over 10 power stations. The total installed capacity is about 40 GW, and nuclear contributes only 6 percent. The existing nuclear power station, Koeberg NPP, is comprised of two 900 MW(e) units at the South African west coast near Cape Town. The Koeberg NPP units are Framatome PWR designs. The South African government has a policy to increase the share of nuclear in the generation mix from 6 percent to 15 percent before 2020. In this regard, the government has approved the design and demonstration of the pebble bed modular reactor (PBMR). The PBMR is a high temperature helium cooled reactor design with a direct cycle. The thermal rating of the reactor is 400 MW(e) with the electrical output of 165 MW(e). The key characteristics of the PBMR design are: - inherent safety, - load following, - modularity and - simple systems

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

    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.

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

    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。

  17. 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

    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% .

  18. CAREM: an innovative-integrated PWR

    Power levels. In this regard the cost effective sizes are 100 MWe (maintaining natural circulation for primary core cooling) and 300 MWe (using integrated primary pumps). (author)

  19. Nuclear safety and environment protection at Cernavoda NPP

    The paper reviews the basic features of nuclear safety of CANDU 600 reactor in operation at Cernavoda NPP and the measures aimed at safety improvement of this type of reactor, especially for the Units 1 and 2. The authors also present the method used for ensuring the environment and population protection during normal operation of NPP as well as in case of emergency. The paper contains the following chapters: 1. Fundamental aspects of nuclear safety at Candu 600 reactor based Cernavoda NPP; 2. A comparison between CANDU-600 and PWR reactors from technological and nuclear safety point of view; 3. Severe accidents with CANDU-600 type reactors; 4. The reactor CANDU-600 facing the European Union requirements for LWR; 5. Design improvements of the CANDU-600 reactor operating at Cernavoda NPP; 6. Environmental and population protection with CANDU-600 reactors and particularly Cernavoda NPP; 7. Emergency plans in the frame of population protection measures

  20. 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

    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 .

  1. Characteristics of Loviisa NPP spent fuel

    The composition and radioactive characteristics of the spent fuel from Imatran Voima Oy's (IVO) Loviisa NPP (VVER-type PWR) have been estimated with the ORIGEN2 computer code (version 2.1) using the so-called PWR-UE cross section library. Four separate cases have been calculated. The main results of the calculations are the composition, activity, heat production, photon source and spectrum, and neutron source of the spent fuel as a function of cooling time. In the tables and figures of the report only the most important data of the large ORIGEN2 output files have been given. The ORIGEN2 results have also been compared with those calculated with the CASMO-HEX fuel assembly burnup program. (11 refs., 3 figs., 10 tabs.)

  2. Mochovce NPP simulator

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

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

    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

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

    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

  5. NPP operation, 2001

    Results of NPP service in 2001 on a global scale are presented. Numerical data on service indexes of NPP in different countries are reviewed. Summary power of operating NPP in 2001 was as much as 372 857 MW. List of ten NPP having the best characteristics in electric power generation on one nuclear bock is given. Nuclear power plants of Germany were recognized as the best units on a global scale

  6. Calculation of behaviour of the Juragua NPP containment with code TRACOV/MOD1

    The containment of Juragua NPP has some unique features, which differ from the rest of the PWR reactors design. Those features impose additional requirements for its numerical simulation. In this paper is analyzed the behaviour of the Juragua NPP containment during accident situation with double ended break of the primary pipelines with flow in both direction using the code TRACOV/MOD1. The results are compared with obtained by the designer. The main restrictions of the code are identified

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

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

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

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

    2015-05-15

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

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

    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)

  10. NPP life management (abstracts)

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

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

    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

  12. Low level radwaste management and processing in Maanshan NPP

    Nuclear power plant like as the other power plant will generate technology waste. Owing to Nuclear still is a debatable topic for discussion, Nuclear radwaste including low level radwaste, high level spent fuel and nuclear operate safety become a focus point in Taiwan also in all world. Maanshan NPP is the only one PWR unit in Taiwan. In common understand, the Low Level radwaste generate from PWR unit is less than BWR. No matter what LLW generate quantity is reduced obviously, the government asks seriously restrain LLW quantity year by year. Maanshan NPP had reach a stable level in solidification waste, system spent resin, combustible and incombustible radwaste that generate from necessary maintenance. The further aim is keep waste generate under control, stable operate processing system and make a new processing technical to dispose spent resin. Maanshan NPP via technical cooperation to set HESS system with INER in one decade. Nowadays there are about 18 55 gallon drums per year in Maanshan NPP. LLW incinerator equipment designed by Maanshan and install at 7 years ago, there almost burns up all the combustible LLW that generate from commercial operation. The new equipment, wet-oxidation solidification process for treatment of spent radioactive ion-exchange resins plan will cooperate with INER and complete in 2014. It is estimated that the generation of solidified wastes of the NPS will be reduced to about 1/3 volume of that currently generated. (author)

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

    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)

  14. On the domestically-made heavy forging for reactor pressure vessels of PWR nuclear power plant

    The present situation of the foreign heavy forgings for nuclear reactor pressure vessels and the heavy forgings condition which is used for the Qinshan 300MWe nuclear power plant are described. Some opinions of domestic products is proposed

  15. Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

    Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

  16. Fragility Assessment of Piping System in Ulchin 5, 6 NPP based on JNES Results

    A Piping system is one of the most important systems in NPP, because a piping system carries coolant of NPP system. Failure of piping system reveals LOCA (loss of coolant accident) which can cause core damage. LOCA divide as large, medium and small LOCA according to a size of piping system. Even though LOCA is one of the most important accidents in NPP, LOCA is only considered in the case of internal event in Korea. But JNES (Japan Nuclear Energy Safety Organization) already performed a fragility analysis about piping systems in PWR and BWR system in Japan. And also Japan considered a failure of piping system in the case of seismic event. In this study, fragility results of Japanese NPP were investigated and fragility of piping system in Korea was evaluated by applying to Japanese method

  17. Economical aspect of the decommissioning for NPP

    The estimated, analysed and founding of the economical aspect at decommissioning of Nuclear Power Plant (NPP) have been studied. The data that have been obtained from literature, then the calculation and analysing have been done base to the future condition. The cost for NPP decommissioning depend on the internal factor such as type, capacity and safe storage time, and the external factor such as policy, manpower and the technology preparation. The successfulness of funding, depend on the rate of inflation, discount rate of interest and the currency fluctuation. For the internal factor, the influence of the type of the reactor (BWR or PWR) to the decommissioning cost is negligible, the big reactor capacity (±1100 MW), and the safe storage between 30 to 100 years are recommended, and for the external factor, specially Indonesia, to meet the future need the ratio of decommissioning cost and capital cost will be lower than in develop countries at the present (10%). The ratio between decommissioning fund and electricity generation cost relatively very low, are more less than 1.79 % for 30 years safe storage, and discount rate of interest 3%, or more less than 0.30 % for safe storage 30 years, and discount rate of interest 6%. (author)

  18. Foreign NPP decommissioning

    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

  19. Zinc addition at ANGRA 2 NPP. A preliminary report

    As a result of an Eletronuclear and Siemens agreement planned to be applied in Angra 2 NPP zinc addition used data from the joint German utilities/Siemens qualification program were as well as operating gathered at the German lead pressurized water reactors plants. The qualification program main objective was to demonstrate the process efficiency, to investigate interactions between zinc and oxide layers, to elaborate a dosing concept, to provide compatibility assessment with systems and components and to develop implementation strategy, defining limiting values and diagnostic parameters and a surveillance program. Angra 2 NPP is the world's first power plant using this program since its start-up in July 15, 2001. Its design features (core design, reactor coolant pumps and others) were also reviewed and compared with corresponding data from German Siemens PWR's, adding zinc. The data showed that the compatibility of method with Angra 2 plant was ensured. (authors)

  20. Survey of Water Chemistry and Corrosion of NPP

    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

  1. Survey of Water Chemistry and Corrosion of NPP

    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.

  2. Assessment of the MSLB accident safety margins for NPP Krsko

    This paper presents the comparison between the 3-D neutronics (coupled code RELAP5/QUABOX/CUBBOX) and point kinetics (RELAP5/mod3.2.2) calculations of MSLB accident for NPP Krsko (NEK) after steam generator (SG) replacement and at uprated conditions (6% power increase). The main steam line break (MSLB) accident in pressurized water reactors (PWR) is overcooling accident characterized by large variations in primary coolant conditions, asymmetric core inlet and outlet conditions and strongly localized reactivity disturbance due to assumed stuck control rod. The characteristics and phenomena related to the MSLB accident for best-estimate calculations can be accurately analyzed through the use of coupled code. (author)

  3. IAEA activities on safety aspects of NPP ageing

    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

  4. EMAS at Doel NPP

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

  5. Bubbling-vacuum installation for accident localization in standard NPP with two WWER-type reactors

    The present installation enchances the efficiency and safety in the elimination of the accident concequencies, as well as the efficiency of the dual-cycle NPP with 2x440 PWR. It is proposed to use only the volume of the usable part of the secondary circuit, e.g., the air chambers. The maximum accident superfluous pressure in the secondary side (linked air chambers) is reduced. The reduction of the superfluous pressure correspond to the reduced total amount of radioactive effluents and to reduction of radioactive burden in tne NPP environment. 2 cls., 2 figs

  6. NPP Krsko decommissioning concept

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

  7. NPP Krsko decommissioning concept

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

  8. Two managerial grids in NPP

    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)

  9. Visualized research on primary loop simulation for PWR nuclear power plant

    In this study the main equipment 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 non- equilibrium 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 equipment 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. (authors)

  10. Brazilian system for NPP personnel training

    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

  11. Plutonium recycling in PWR

    Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)

  12. The integrated PWR

    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.)

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

    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)

  14. Reconsidering the site requirements for NPP on Olt River

    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

  15. PWR type reactor

    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.)

  16. PWR decontamination feasibility study

    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

  17. Control of NPP aging processes

    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

  18. The Influence of Using Moisture Separation to 1000 MWe PWR Typed Nuclear Power Plant Performance

    Thermodynamics analysis for to know influence using moisture separation to efficiency upgrading of 1000 MWe Pressurized Water Reactor typed Nuclear Power Plant (NPP) performance have been done. The thermodynamic analysis to power plant for use performance planning and predicting, with the result that not over of operational condition limited.The Pressurized Water Reactor typed of NPP using Rankine Cycle in the thermodynamic analysis can be upgrade by using moisture separation. The function of moisture separation is separate of some water which formed from turbine on high pressure area and then to be fill up in one stage of feedwater heating (OFWH) for then to circulation in the primary system. The resulted from analysis show that using moisture separation can be increase efficiency PWR 1000 MW(e) typed of NPP from 30.4 % to 37.3 % or increase 6.9 %. (author)

  19. Test research and analysis for ultimate capacity of Qinshan NPP PCCV

    This paper introduces design and research for containment of Qinshan NPP which is the first PWR in CHINA designed and constructed by ourselves. The PCCV design is basically in conformity to ASME code. To verify the structural integrity capacity of Qinshan NPP containment, we fulfilled SIT and ILRT successfully in June, 1991. The special attention of the paper is focused on the ultimate capacity of the PCCV under severe accidents and earthquake. A study comprised of five different independent parts has been performed for the development of containment model test and corresponding nonlinear analysis. There are two prestressed concrete containment models with equipment hatch. One is 1/15 scale with steel liner tested on shake table and then moved out loaded with atmospheric pressure. The other is 1/10 scale without steel liner loaded with water pressure until destruction. From different methods including model test and nonlinear analysis, all obtained unanimous conclusion. The capacity under internal pressure and earthquake is reliable. The safety margin is enough. Consequently, in the second phase of Qinshan NPP and other PWR NPP under design, PCCV should be a better selection in China since it's more economic, rational and safe. (author)

  20. Russian NPP I and C systems and NPP safety problems

    The long experience of nuclear power plant (NPP) operation both in Russia and over the world confirms that both power and economic characteristics as well as NPP safety depend on possibilities and specifications of instrumentation and control (I and C) systems. That is why the more serious attention is paid to the problems of improvement of I and C systems in all countries

  1. NPP Evaluation, backfitting and life extension. An engineering viewpoint

    During the decade of the 80s, the Owners of the two oldest operating plants in Spain designed and built during the 60s - namely, Jose Cabrera NPP, a Westinghouse PWR, and Santa Maria de Garona NPP, a GE BWR- undertook the following important programs: 1. A far-reaching Systematic Evaluation Program (SEP) for the Jose Cabrera NPP consisting in the systematic safety review of the plant design, followed by the necessary hardware modifications, to upgrade it and make it comply with current safety criteria, and a Plant Upgrading Program for the Garona Nuclear Station focusing on specific topics affecting GE BWR Mark-I type plants of the same vintage. 2. A Remaining Life Management Program to ensure that the units, after extensive backfittings and high capital investment, would complete their design life, leaving open the option for plant life extension. These two units are today considered by the Spanish nuclear industry as the pilot plants for Plant Life Extension (PLEX) programs for PWRs and BWRs in our country The purpose of this paper is to summarize the principal lessons learned from EMPRESARIOS AGRUPADOS' participation as an architect-engineering organization in the engineering, design and implementation of these Programs. They are practical examples of positive experience which could be considered as a reference when carrying out similar programs for other plants. (author)

  2. Preparation for Ignalina NPP decommissioning

    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

  3. IAEA recommendations on NPP safety

    Codes developed in IAEA on the basis of the NUSS program (Nuclear Safety Standards) concerning nuclear safety of thermal reactor NPPs and published in 1978 are considered. 5 main codes and manuals have been stated: 1. Governmental organization for the regulation of NPP; 2. Safety in NPP siting; 3. Design for safety of NPP; 4. Safety in NPP operation; 5. Quality assuarance for safety in Nuclear Power Plants. The Codes contain recommendations on providing safety of population and personnel as well as on environmental protection. They also contain criteria and proper measures corresponding to both operating conditions of NPP and possible emergency conditions. Some provisions in the Codes may be also used, for providing radiation safety and at the external fuel cycle plants

  4. PWR degraded core analysis

    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.)

  5. Full System Decontamination (FSD) with the CORDR Family prior to Decommissioning - Experiences at the German NPP Obrigheim 2007

    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 m3, total system surface approximately 8100 m2. 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)

  6. Automated personal dosimetry monitoring system for NPP

    Chanyshev, E.; Chechyotkin, N.; Kondratev, A.; Plyshevskaya, D. [Design Bureau ' Promengineering' , Moscow (Russian Federation)

    2006-07-01

    Full text: Radiation safety of personnel at nuclear power plants (NPP) is a priority aim. Degree of radiation exposure of personnel is defined by many factors: NPP design, operation of equipment, organizational management of radiation hazardous works and, certainly, safety culture of every employee. Automated Personal Dosimetry Monitoring System (A.P.D.M.S.) is applied at all nuclear power plants nowadays in Russia to eliminate the possibility of occupational radiation exposure beyond regulated level under different modes of NPP operation. A.P.D.M.S. provides individual radiation dose registration. In the paper the efforts of Design Bureau 'Promengineering' in construction of software and hardware complex of A.P.D.M.S. (S.H.W. A.P.D.M.S.) for NPP with PWR are presented. The developed complex is intended to automatize activities of radiation safety department when caring out individual dosimetry control. The complex covers all main processes concerning individual monitoring of external and internal radiation exposure as well as dose recording, management, and planning. S.H.W. A.P.D.M.S. is a multi-purpose system which software was designed on the modular approach. This approach presumes modification and extension of software using new components (modules) without changes in other components. Such structure makes the system flexible and allows modifying it in case of implementation a new radiation safety requirements and extending the scope of dosimetry monitoring. That gives the possibility to include with time new kinds of dosimetry control for Russian NPP in compliance with IAEA recommendations, for instance, control of the equivalent dose rate to the skin and the equivalent dose rate to the lens of the eye S.H.W. A.P.D.M.S. provides dosimetry control as follows: Current monitoring of external radiation exposure: - Gamma radiation dose measurement using radio-photoluminescent personal dosimeters. - Neutron radiation dose measurement using

  7. Automated personal dosimetry monitoring system for NPP

    Full text: Radiation safety of personnel at nuclear power plants (NPP) is a priority aim. Degree of radiation exposure of personnel is defined by many factors: NPP design, operation of equipment, organizational management of radiation hazardous works and, certainly, safety culture of every employee. Automated Personal Dosimetry Monitoring System (A.P.D.M.S.) is applied at all nuclear power plants nowadays in Russia to eliminate the possibility of occupational radiation exposure beyond regulated level under different modes of NPP operation. A.P.D.M.S. provides individual radiation dose registration. In the paper the efforts of Design Bureau 'Promengineering' in construction of software and hardware complex of A.P.D.M.S. (S.H.W. A.P.D.M.S.) for NPP with PWR are presented. The developed complex is intended to automatize activities of radiation safety department when caring out individual dosimetry control. The complex covers all main processes concerning individual monitoring of external and internal radiation exposure as well as dose recording, management, and planning. S.H.W. A.P.D.M.S. is a multi-purpose system which software was designed on the modular approach. This approach presumes modification and extension of software using new components (modules) without changes in other components. Such structure makes the system flexible and allows modifying it in case of implementation a new radiation safety requirements and extending the scope of dosimetry monitoring. That gives the possibility to include with time new kinds of dosimetry control for Russian NPP in compliance with IAEA recommendations, for instance, control of the equivalent dose rate to the skin and the equivalent dose rate to the lens of the eye S.H.W. A.P.D.M.S. provides dosimetry control as follows: Current monitoring of external radiation exposure: - Gamma radiation dose measurement using radio-photoluminescent personal dosimeters. - Neutron radiation dose measurement using thermoluminescent

  8. NPP Prevlaka - Preparation of construction

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

  9. The seismic reassessment Mochovce NPP

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

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

    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)

  11. The PWR programme

    For fueling the PWR type reactors two types of fuel were developed: the UO2 and mixed oxide fuels. To satisfy the demand of the operators of UO2-fuelled power plants a specific industrial organization has been established by Cogema and Framatome: Framagema supplies the technical expertise and sells the fuel; FBFC (Societe Franco-Belge de Fabrication Combustible) is manufacturing the fuel by using particularly the zirconium components produced by Zircotube and Cezus. By making possible the recycling of the materials recovered from the spent fuel reprocessing the MOX (mixed oxide fuels) technology represents an important venture for the future electronuclear sector. To implement this project Cogema created together with Belgonucleaire (the administrator of the Dessel manufacture plant) the GIE COMMOX, in charge with marketing of this fuel. On the other side Cogema which produces MOX in its facility at Cadarache, is at present building the plant at Melox of a capacity of 120 tonnes/year. After presenting the present situation with UO2 and MOX fuels the paper ends with considerations concerning the future fuels and fuels for future and further future reactors

  12. 'Independent' monitoring of the aerosol effluents from NPP provided by SURO

    In this paper there are the results of the independent monitoring of the aerosol discharges into the environment from NPPs of Dukovany and Temelin, which SURO provided for many years. The aerosol discharges are followed in 2 ventilation stacks of the Dukovany NPP (each ventilation stack is common for 2 pressurised water reactors of PWR Novovoronezh, type 213, 440 MWe) and in 5 ventilation stacks of the Temelin NPP (2 pressurised water reactors of PWR Novovoronezh, type VVER-1000/320, 1000 MWe, each of them connected to the own inner ventilation stack and outer ventilation stack, which surrounds the inner one; 1 ventilation stack is for the building of active and instrumental services). A system of the pipelines, whereby a sample of the exhausted air with aerosols is led away for different analyses, begins at a height of 40 - 50 m above the bottom of the ventilation stack. The pipelines system ensures the isokinetic sampling. Besides independent monitoring of the aerosols SURO provides the independent monitoring of the noble gases including 85Kr and 14C in the effluents from the ventilation stacks, too. This monitoring isn't performed continuously, because sampling and monitoring are relatively demanding. SURO also performs over the frame of the independent monitoring the determination of the aerosol size distribution from the Temelin NPP in dependence on the kind of a radionuclide. These determinations are in the monitoring plan of the Temelin NPP, but they aren't quite routine. The similar measurements took place in Dukovany NPP (VK-1) in the time period 1999-2001. The results can enter to the models of radionuclide diffusion in the environment and they are also useful for research. The 'independent' monitoring of the NPPs is a significant part of the monitoring of the Czech Republic territory, it confirms the values presented by operator and provides also a valuable base for research tasks. (authors)

  13. Selection of NPP for Kazakhstan

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

  14. Analyses and estimation of insulation material release in E.ON-PWR under loss of coolant conditions

    In 1992, strainers on the suction side of the ECCS pumps in Barsebaeck NPP Unit 2 became partially and temporary clogged with mineral wool after steam jet induced releases of parts of the mineral wool insulation. Although Barsebaeck NPP Unit 2 is a Boiling Water Reactor (BWR) this event induced large investigations to understand and maintain strainer clogging effects after a loss-of-coolant accident for all reactor types. Especially for the German Pressurized Water Reactors (PWR) a program was launched by the German Utilities together with the plant manufacturer. The final estimations of all theoretical and experimental results caused different steps of modifications of the insulation material, the strainer area and mesh sizes in most of the German PWR including the E.ON PWR. Moreover, interactions and procedures were carried out such as back flushing actions to remove the mineral wool from the strainers supported by an additional differential pressure measurement at the strainers. All measures were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode and to improve the NPP safety. (orig.)

  15. Psychology of NPP operation safety

    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

  16. To question of NPP power reactor choice for Kazakhstan

    Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

  17. To question of NPP power reactor choice for Kazakhstan

    The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

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

    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

  19. Sizewell 'B' PWR reference design

    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.)

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

    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)

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

    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)

  2. Safety Review models on radioactive source term design for PWR waste treatment systems

    The source of all liquid, gaseous and solid radioactive waste in the pressure water reactor (PWR) nuclear power plant (NPP) originate from leakage of fission products out of the fuel rods into the primary coolant and neutron activation of materials within and around the primary coolant system and reactor vessel. The source term design used to determine the concentrations of radionuclides in the reactor coolant, which could be: (1) A conservative source term which predicts the maximum concentration of nuclides in the Reactor Coolant System establishes the design basis of the various onsite processing systems for the purpose of defining system capacity and shielding requirements and (2) A realistic source term for the purpose of evaluating the reasonably expected inventories and releases of radionuclides under normal operating condition, including anticipated operational occurrences. This paper will discuss the safety review models on source term design for the PWR waste system mainly on the basis of the conception of the conservative source term. (author)

  3. Development and validation of the 3-D PWR core dynamics SIMTRAN code

    We discuss the main features and results of the SIMTRAN development and validation work. Included in the first are the extension of the nodal neutronic solution to account for intranodal shape and spectrum, due to both heterogeneities and flux gradients, the implicit scheme for spatial kinetics with six delayed neutron precursors and the integration of the neutronic and thermohydraulic solutions on an staggered time mesh. Validation results are discussed for the NEACRP 3-D PWR Core Transient Benchmark and an actual transient with sudden increase of core flow occurred in the Vandellos-II 3-loop PWR NPP. Agreement with the reference numerical solution and measured plant data is shown for both problems. (orig./DG)

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

    Experiments are being conducted on nuclear power plant (NPP) control room training simulators by the Oak Ridge National Laboratory, its subcontractor, General Physics Corporation, and participating utilities. The experiments are sponsored by the Nuclear Regulatory Commission's (NRC) Human Factors and Safeguards Branch, Division of Risk Analysis and Operations, and are a continuation of prior research using simulators, supported by field data collection, to provide a technical basis for NRC human factors regulatory issues concerned with the operational safety of nuclear power plants. 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 (a) senior reactor operator (SRO) experience, (b) operating crew augmentation with an STA and (c) practice, as a crew, upon crew and individual operator performance, in response to anticipated plant transients. The FY84 experiments are a partial replication and extension of the FY83 experiment, but with PWR operators and simulator. Methodology and results to date are reported

  5. Temelin NPP status and challenge

    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

  6. Emergency preparedness at Ignalina NPP

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

  7. Teledosimetry system of Mochovce NPP

    On-site monitoring posts, inner circuit there are installed 16 detectors for measuring dose rates in a circle line inside the power plant site at the distances between approximately 200 m and 520 m from the exhaust air stack of Mochovce-1,2 NPP. The technical versions of all these measuring posts are from the same type On-site are installed 3 large containers for measuring gamma dose rate and activity concentration of aerosols and iodine too. Off-site monitoring posts at places of living consist of 16 small containers with dose rate measurements and iodine sampling units and 8 large containers with dose rate measurement and aerosol and iodine monitoring unit. Three of these large containers are installed nearby the NPP inside the fence of the NPP area. The technical version of these measuring posts are from the same type. Five of these containers are installed faraway of the NPP outside the fence of the NPP. These five posts are from the same type. The process data is continuously acquired, stored and processed by the Central Radiological Computer System of the power plant. The tele-dosimetry system data are a main part of the radiological information system, which continuously provide the information of the measurements and evaluate possible radiological consequences. (author)

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

    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)

  9. Velocity of crack growing of Inconel-600, sensitized, contaminated with sulphur in PWR type reactors

    The origin of the vessel head penetration cracking of Jose Cabrera NPP has been attributed to an IGA/SCC process in a highly sensitized Alloy 600 assisted by sulphur species, as both acid sulphates and reduced species originated by the thermal breakdown of the cationic resins present in the primary coolant. The thermal degradation of the cationic resins leads sulphonic acid group scission and sulphates. Under the operating conditions the reduction of sulphates to sulphides is produced. The sulphides formed from the reduction of sulphate can precipitate with metallic cations and be incorporated into the oxide layers of the materials, preferably into nickel alloys. Others components at Jose Cabrera NPP are fabricated from sensitized alloy 600, as bottom vessel penetrations. In order to determine the influence of sulphur incorporated to the oxide layers of bottom vessel penetration alloy 600, an experimental work has been performed to obtained crack growth rate data under PWR primary conditions on sensitized alloy 600. (Author) 5 refs

  10. PWR AXIAL BURNUP PROFILE ANALYSIS

    J.M. Acaglione

    2003-09-17

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

  11. Condensate purification in PWR reactors

    The recommendations made by the VGB task group on 'condensate purification for PWR reactors' 1976 are discussed in detail. Techniques and circuiting possibilities of condensate purification for BBR steam generators (forced circulation) and KWU steam generators (U tube with blow-down) are mentioned. (HP)

  12. PWR AXIAL BURNUP PROFILE ANALYSIS

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

  13. The conceptual design of IPR1000 reactor pressure vessel for PWR type

    The conceptual design of IPR1000 reactor pressure vessel for PWR type has been configured, material selection and dimension parameter are designed to cover the reactor cooling system (RCS), nuclear fuel assembly, and others internals reactor. The reactor pressure vessel consist of closure head assembly, vessel shell upper head assembly, vessel shell lower assembly, and inlet and outlet nozzle. These are designed capable to support weight of RPV, at pressures and temperature of each 2485 psig and 650 °F. The design refers to AP1000 as according to ASME code and industrials standard applicable for Nuclear Power Plants (NPP). (author)

  14. Primary analysis of PWR loaded with MOX fuel and related fuel cycle scenarios in China

    To meet the China's energy demand, nuclear power will keep growing in the future. Nuclear fuel cycle system is essential for the nuclear power development in China. In this paper, nuclear fuel cycle issues, including the amount of natural uranium resource, separation work and nuclear fuel for PWR NPP, together with spent fuel and separated plutonium are studied. The influences of spent fuel reprocessing and separated plutonium recycling on the uranium resource demand and accumulation are discussed in two fuel cycle scenarios. (authors)

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

    Thunem, Harald P-J; Zhang, Xinxin

    2015-01-01

    This paper reports on the results from the practical application of the Shape Shifter framework on the continued development of a graphical editing suite, the MFM Suite, for MFM and process model design and analysis. The primary use of the MFM Suite is diagnosis and prognosis of anomalies in 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....

  16. Ignalina NPP Safety Improvement Program

    After 1991, when Lithuania became independent, the new safety improvement activities were initiated. Lithuania committed its responsibility for Ignalina NPP safety. However, it was lack of money for adequate safety improvement. Countries of Western Europe, USA and IAEA assisted Lithuania to carry out a comprehensive Ignalina NPP safety improvement program. Agreement with EBRD was signed in 1994. As a result of some bilateral and multilateral cooperation projects the Ignalina NPP Safety Improvement Programme (SIP-1) was accepted in 1993.This program was being implemented during 1993-1996. The Safety Analysis Report was issued in 1996. Review of the SAR was performed and RSR report was issued. On basis of both documents the Ignalina Safety Panel prepared recommendations for Lithuanian Government. These documents were used as a basis for the Safety Improvement Programme No.2. SIP-2 was accepted in 1997 and shall be finished in 2000

  17. Safety culture at Mochovce NPP

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

  18. Regulatory aspects of NPP safety

    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

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

    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

  20. Supervision of Operation Safety at Ignalina NPP

    Description of VATESI supervision functions during operation and maintenance of Ignalina NPP is presented. Comparison of collective exposure dose of Ignalina NPP and other organizations with previous years is made. The number of emergency outings of Ignalina NPP units during the years is presented

  1. Training of experts on NPP decommissioning

    The paper presents difficulties and problems in training of NPP decommissioning experts in Ukraine. The scientific and technical cluster is offered to be constructed as a territorial association of enterprises and organizations related to NPP decommissioning issues and spent nuclear fuel and radioactive waste management. The center is to be based on scientific and educational center in Slavutych, satellite city of Chornobyl NPP.

  2. NPP Krsko secondary side analysis

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

  3. Radioactive wastes management of NPP

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

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

    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. RCM at Kozloduy NPP, Bulgaria

    RCM methodology is applied within the task for Maintenance and Repair Optimization which is part of a big project started in 2005 for Optimization of maintenance using risk-informed PSA applications for Units 5 and 6 (VVER 10001320) of Kozloduy NPP. The project is still under development

  6. Overview of PWR chemistry options

    Nordmann, F.; Stutzmann, A.; Bretelle, J.L. [Electricite de France, Central Labs. (France)

    2002-07-01

    EDF Central Laboratories, in charge of engineering in chemistry, of defining the chemistry specifications and studying the operation feedback and improvement for 58 PWR units, have the opportunity to evaluate many options of operation developed and applied all around the world. Thanks to these international relationships and to the benefit of a large feedback from many units, some general evaluation of the various options is discussed in this paper. (authors)

  7. PWR secondary water chemistry guidelines: Revision 3

    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)

  8. Risk indicators at Cofrentes NPP

    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. Analysis of a rod withdrawal in a PWR core with the neutronic- thermalhydraulic coupled code RELAP/PARCS and RELAP/VALKIN

    The Reactor Ejection Accident (REA) belongs to the Reactor Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressure water reactors (PWR). The REA at hot zero power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS2.5 and RELAP5/VALKIN a REA in Trillo NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions. (authors)

  10. Safety enhancement at Beznau NPP

    The two units of the Beznau Nuclear Power Plant, Switzerland, are presented, and their safety related progress is evaluated. The largest safety enhancement has been the addition of a completely self-contained emergency system. Safety enhancements through backfitting measures in older nuclear power plants, however, have distinctive disadvantages compared to more modern plants. At Beznau NPP, safety has always priority over economics. (N.T.)

  11. Artificial intelligence and NPP safety

    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

  12. New approach of second Romanian NPP siting

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

  13. Experiences with Simulator Training of NPP Staff in Germany

    Simulator training for NPP staff has a long history in Germany. Our first full scope simulator was taken into operation in 1977. Since then the simulator training for almost all German plants as well as for the Dutch NPP Borssele and until a few years ago for the Swiss NPP Goesgen-Daeniken was performed in the simulator center in Essen. Since the plants are technically rather different, a high number of plant specific simulators is needed to cover their training needs. To serve the 6 BWR and 14 PWR units in operation in Germany and the Netherlands, 13 full scope simulators are operated at the training center and one at the NPP Kruemmel. The simulators are specific to one unit or a few technically similar ones (e.g. 3 Konvoi plants). They have a high degree of fidelity, especially the IC of the highly automated plants is simulated to a great detail. The simulator center performs about 500 one week courses a year. Most of them are for shift personnel, a smaller part for NPP management and a few for other personnel. About one third of the courses for the shift personnel are initial courses, about two thirds are retraining courses. The training which the center provides today reflects our experience during the last more than twenty years. Concepts and guidelines were developed for program development, course preparation, supporting documents, trainee assessment, instructor training, etc. The goal is to provide training for all plants as far as possible according to the same standards and design it at the same time as plant specific as possible. For the latter purpose the training is developed in close cooperation with the training departments of the plants. Plant managers observe parts of practically all courses. Until about five years ago accidents with the failure of safety systems beyond the minimum design requirements played a minor role in the training. About at this time the NPP's had systematically developed procedures for accident management measures for

  14. Thermodynamic modelling of PWR coolant

    Spinel solubilities on PWR primary circuit surfaces vary with temperature, pH and coolant H2 concentration. The available solubility data are discussed for Fe, Ni, Co and Zn oxides, and species are identified where data are very limited or absent. An equilibrium thermodynamic model is described to predict the solubility, and results are described predicting relative Fe and Ni solubility under normal operating conditions and during shutdown/startup. The relative stabilities of stoichiometric and non-stoichiometric zinc ferrite spinels are also considered. (R.P.)

  15. Effective long term operation for Dukovany NPP

    Dukovany NPP now started third decade of service that is also its last decade of design life time. It is clear that the NPP has all considerations for service past the design life time called Long Term Operation (LTO). This LTO has two main aspects, technical and economical, that influence each other. From technical view the age of NPP systems, structures and components (SSCs) affects negatively the ability to perform necessary design changes in a good quality and also the long lived SSC reliability. These possible impacts have also their safety aspects and to obtain regulatory body agreement with LTO of NPP it is necessary to show that these impacts are acceptable. It means to show that all applied design changes are done in agreement with NPP design bases (DB) and all ageing impacts on SSCs functions important for safety are properly managed. From economical view that is significant for NPP owner it is necessary to demonstrate a required profitability of investment for effective LTO. These are reasons why Dukovany NPP performs three following projects: - Safety design bases collation and reconstitution, - Enhancement of plant life management program (New program preparation), - Technical-economical (TE) study of NPP LTO. All of these projects are managed by Nuclear Research Institute Rez plc (NRI) and performed in close cooperation with NPP staff and different co-operaters. This presentation will be concentrated to the last named project.

  16. NPP technical and economical parameters and safety

    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

  17. Ignalina NPP: living and working conditions

    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

  18. Metamorphosis of NPP A1, V1, V2

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

  19. Reliability analysis of 2 types of auxiliary feedwater system for PWR

    This paper will explain the application of Fault Three Method for analyzing the system reliability of Auxiliary Feedwater System with 2 different configurations taken from PWR type nuclear power plant (NPP) in the USA. The first configuration of Braidwood NPP (design A) basically consists of 1 motor driven pump and 1 diesel driven pump. The second configuration of Haddam Neck NPP (Design B) consists of 2 turbine driven pumps. Based on the P and ID and success criteria the fault trees are constructed to estimate the system failure probabilities quantified from software code PIRAS 1.0. The result shows the second configuration (Design B) with 2 turbine driven pumps have the higher failure probability of 1,06 x 10 -2 compared with design A of 1,09 x 10-3. The modification of both systems are also tried to analyze its effect to the end result. Qualitatively, the common cause failures of 2 turbine driven pumps contribute to the highest risk of system failure probability. Combination with 1 turbine driven pump and 1 motor driven pump or 1 diesel driven pump will increase the system reliability about 80% and 50% without considering if this configuration is possible to realize in a real plant

  20. Conceptual design of simplified PWR

    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

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

    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

  2. Chernobyl NPP accident. Overcoming experience. Acquired lessons

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

  3. Emergency preparedness at Barsebaeck NPP in Sweden

    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

  4. Psychological methods as applied to NPP personnel

    Psychologists' experience in nuclear power personnel work system is described. Possibilities of practical application of scientific information, ways and methods collected in psychology, their effect when solving problems on profession orientation, personnel selection, arrangement, training and education, are shown. Necessity to take into account personnel psychological data under conditions of increased hazard of work at NPP is illustrated taking Chernobyl NPP as an example

  5. Safety culture in Ignalina NPP, regulatory view

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

  6. VMEbus technology in NPP control automation

    In frames of SPAS project (system for emergency situations and accident prevention at NPPs in Ukraine) a series of developments was made to increase the efficiency and control of NPP equipment and main technological processes. They are based on information which is permanently renewed and accumulated in regular NPP system. Technical parameters of this system are described

  7. Seismic characterization of the NPP Krsko site

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

  8. Technical support to an operating PWR vis-a-vis safety analysis

    Currently a PWR of 300 MWe capacity CHASNUPP-I is in operation since the year 2000. Technical support being provided includes In-core fuel management and corresponding safety analysis for the reshuffled core for the next cycle. Prior to start of cycle six an extension in cycle five based on coast down technique was achieved of almost 30 effective full power days. Cycle 6 was designed to achieve the safe and economical loading pattern. The technique used is designated as out-in mode (modified). In this technique, most of the fresh fuel assemblies are not directly located at the periphery of the core, but near the boundary. This technique has the advantage that no burnable absorbers are used in each cycle and we get less radial neutron leakage and increased discharge burnup and cycle length. Operating experience/feedback shows that this type of loading pattern gives better economy without resorting to the conventional in-out technique. The lifetime of the cycle is predicted as 10371 MWD/MTU or 373 Effective Full Power Days (EFPD at 998.6 MWth). In design calculations, the end of cycle is reached at 10 ppm critical boron concentration in the unroded core. Measured critical boron concentration at HZP, BOL is 1453 ppm compared with the calculated value i.e 1457 ppm, is within the acceptable limits. It is also observed that the calculated reactivity worth of Tl is -1771 pcm as compared to measured value i.e -1802 pcm with difference of only 1.6 % showing the reliability of the design value. The measured Moderator temperature coefficient (MTC) is 2.52 pcm/deg. C at all rods out (ARO) and critical boron concentration (CBC) condition whereas the calculated value is 3.36 pcm/deg. C (at predicted CBC of 1457) having a good agreement with design value. Safety evaluation of cycle 6 was carried out for the reshuffled core. All the probable accident scenarios based on initiating events as given in the FSAR were evaluated with respect to input parameters. For a specific event, the

  9. PWR fuel: experience and development

    The start-up of the large French nuclear program has rapidly led FRAGEMA to be one of the first PWR fuel suppliers. FRAGEMA is a joint subsidiary of two companies whose scopes of supply are fully complementary: FRAMATOME (NSSS vendor) and COGEMA (nuclear fuel cycle service supplier). At the center of these two activities FRAGEMA is in charge of designing and marketing fuel assemblies. Assistance is also offered to nuclear power plant operators in all fuel related fields by providing a wide range of services and a number of specialized components. Over the past years a statistical data base has been accumulated on fuel assembly behaviour under various operating conditions. At the same time extensive experimental programs have been, set up to develop advanced products to cope with utilities needs in the future. An overview of these two sides of our experience is presented in the following

  10. PWR standardization: The French experience

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  11. [Methodologies for optimization of maintenance and testing of safety related equipment at NPPs in Pakistan

    In Pakistan, a 137 MWe PHWR type NPP (KANUPP) is in operation since 1971, and a 300 MWe Chinese design PWR (CHASNUPP) is under construction. The under construction PWR is planned to be connected to the national grid in 1998. Under this Coordinated Research Project, the work is planned to be carried out for improvement and optimization of the maintenance and surveillance programme for safety related systems and equipment of the above mentioned two NPPs. Efforts will be directed to acquire latest knowledge regarding various methods and strategies for surveillance testing and plant technical specifications through exchange of information. This project will provide a good opportunity to the regulatory body regarding development of acceptance criteria for testing and maintenance of safety related systems and equipment. 8 refs

  12. Seismic risk at NPP Cernavoda

    The paper presents a brief description of the probabilistic analysis method employed to analyze a nuclear power plant (NPP) state during seismic events. For evaluating the seismic hazard at NPP Cernavoda, deterministic judgment are employed for determining the focus position, the attenuation of the seismic intensity and the local effects. The authors have designed a support device which can operate in damping regimes or as a snubber having also the function of non-linear elastic support with a progressive characteristic for permanent loads. This support can take over very great static and dynamics loads. The support and damping devices are hydro-pneumatic systems which completely eliminate the gaskets. The non-linear elastic supporting force is provided by compressing a volume of gas and the damping force is provided by a liquid flowing through a system of calibrated nozzles covered on one side or the other by a set of elastic blades of an imposed stiffness. The sizes and weight of the devices are under the ones of the current devices and they provide higher reliability. (Author)

  13. Operating Experience at NPP Krsko

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

  14. Reviewing NPP Cernavoda site evaluation

    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)

  15. The decommissioning NPP A-1

    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

  16. Safety culture at Ignalina NPP

    In accordance with Article 27 of the Law on Nuclear Energy of the Republic of Lithuania, the organization operating the nuclear power facility must ensure adequate safety culture. Safety culture comprises specific features and characteristics of the organisation's activities as well as human behavior ensuring that the issues of a nuclear power facility's safety will be given attention consistent with their importance. To ensure adequate level of safety culture, Ignalina NPP has been following IAEA recommendations. The INPP draws up and implements plans of safety culture assurance every year. The Director General meets with IAEA personnel on a regular basis and discusses issues that hold most interest for them. In 2005, INPP management reviewed and approved on September 30 the new policy of safety and quality assurance. The document differs from the policy approved in 1995 in that priorities are set for INPP decommissioning. It is emphasized that INPP Unit 2 operation must be terminated in the most efficient and safest manner, with adequate social security of the personnel assured and effective management system of the facility maintained. The work commenced in 2004 at the Ignalina NPP on identification and application of safety culture indicators was continued in 2005. (author)

  17. The Tokai NPP decommissioning technique

    Tokai power station was closed down in March 1998 and started decommissioning from December 2001 as a pioneer of NPP decommissioning. This article presented current state of Tokai NPP decommissioning technique. As the second stage of decommissioning works, removal works of steam raising unit (four units of heat exchangers) were started from 2006 by jacking down method with decommissioning data accumulated. Each heat exchanger was divided into top head, seven 'tears' of shell and bottom head. Each 'tear' was out and separated into a cylinder, and then divided into two by remote-operated cutting equipment with manipulators for gas cutting and motor disk cutting under monitoring works by fixed and mobile cameras. Divided 'tear' was further cut into center baffle plate, heat transfer tubes and fine pieces of shell. Cutting works would produce radioactive fine particles, which were filtered by temporary ventilation equipment with exhaust fan and filters. Appropriate works using existing technique combined and their rationalization were important at this stage. (T. Tanaka)

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

    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

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

    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)

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

    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.

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

    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

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

    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.

  3. Design and construction of the PCPV for the 300 MWe THTR nuclear power station in West Germany

    In July 1972 the order was placed in Germany for the first PCPV comprising concrete structure, liner, cooling system and insulation to a consortium under the direction of KRUPP UNIVERSALBAU. The prestressed concrete structure itself was designed and constructed by this company. Extensive tests were carried out on the limestone concrete to establish all the physical properties. Special efforts were made to produce a mix which was both pumpable and generated a minimum amount of heat of hydration. As a departure from normal practice, the cylindrical parts of the vessel are constructed in complete rings up to 2m in height and of the full wall thickness. Experiments showed that, for this method of construction, the temperature difference between the old and the new concrete should not be allowed to exceed 50C. To achieve this, ice cooled water is used in the concrete mix and, in the summer time, liquid nitrogen is added at the time of mixing. The thermal behavior of the concrete has been monitored throughout the construction period. A novel construction feature worth mentioning is that the internal insulation and parts of the core structure were already erected before the construction of the concrete cylinder was complete. This was achieved by providing a temporary closure at the top of the cylinder to maintain clean conditions below. The overall stress calculations and the detailed stress pattern for the lower half of the vessel were carried out by using an axi-symmetric computer program but, for the upper half of the cylinder, a three-dimensional analysis was necessary (due to its geometric arrangement). To prove the safety of the vessel a structural model was used from which the mode of failure was found using a kinematic chain and thus the factor of safety established. A secondary line of safety is the integrity of the liner. (author)

  4. Radioecological problems of NPP reservoirs-coolers

    Radioecological problems of NPP reservoir-coolers are considered in connection with thermal effluents and partly radiactive wastes. It is shown that one of real means to reduce undesirable ecological consequences of surplus heat release into the medium is the usage of NPP heated waters in energy-biological and agro-industrial complexes. In case of NPP operation the normalized environmental disposal of a number of radionuclides is specified. In this connection the necessity is pointed out to establish a list of the most dangerous radionuclides to be discharged into water medium by various NPP types, to study their behaviour in main water reservoir components; to determine coefficients of radionuclides accumulation in organisms related to human food chain. Actual is the problem of biological effects which can arise in hydrocenoses of reservoir-coolers as a result of long-term or chronic action of NPP radioactive waste disposal. A wide program of ecological investigations is laid down related to the problem of using NPP water thermal effluents and radioecology of reservoirs-coolers, the realization of the program being initiated in the vicinity of the Beloyarsk NPP

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

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

  6. Improved technical specifications for Korean NPP

    PWRs use Technical Specifications(Tech. Spec.) to ensure safe operation of the plant. Recently, many efforts were made to improve Tech. Spec. and as a result, Improved Standard Technical Specifications(ISTS) have been developed. Korean NPP technical specifications were converted to ISTS format. KAERI also provided supporting documents for technical specification conversion including mark-up's and description of changes. This paper describes and summarizes the results of implementation of ISTS for Korean NPP. The new Tech. Spec. will improve safety of Korean NPP

  7. A brief overview of Ignalina NPP safety issues

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

  8. KEPIC application on Korean NPP

    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

  9. Fuel reliability of Bohunice NPP

    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)

  10. Suomi NPP VIIRS Imagery evaluation

    Hillger, Donald; Seaman, Curtis; Liang, Calvin; Miller, Steven; Lindsey, Daniel; Kopp, Thomas

    2014-06-01

    The Visible Infrared Imaging Radiometer Suite (VIIRS) combines the best aspects of both civilian and military heritage instrumentation. VIIRS has improved capabilities over its predecessors: a wider swath width and much higher spatial resolution at swath edge. The VIIRS day-night band (DNB) is sensitive to very low levels of visible light and is capable of detecting low clouds, land surface features, and sea ice at night, in addition to light emissions from both man-made and natural sources. Imagery from the Suomi National Polar-orbiting Partnership (Suomi NPP) satellite has been in the checkout process since its launch on 28 October 2011. The ongoing evaluation of VIIRS Imagery helped resolve several imagery-related issues, including missing radiance measurements. In particular, near-constant contrast imagery, derived from the DNB, had a large number of issues to overcome, including numerous missing or blank-fill images and a stray light leakage problem that was only recently resolved via software fixes. In spite of various sensor issues, the VIIRS DNB has added tremendous operational and research value to Suomi NPP. Remarkably, it has been discovered to be sensitive enough to identify clouds even in very low light new moon conditions, using reflected light from the Earth's airglow layer. Impressive examples of the multispectral imaging capabilities are shown to demonstrate its applications for a wide range of operational users. Future members of the Joint Polar Satellite System constellation will also carry and extend the use of VIIRS. Imagery evaluation will continue with these satellites to ensure the quality of imagery for end users.

  11. Krsko NPP radioactive waste characteristics

    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

  12. Maintenance robot for PWR plant

    The remote operation, automatic machines utilized in the field of the maintenance of component machinery and equipment in nuclear power plants, so-called maintenance robots, have produced effects in the reduction of radiation exposure, the improvement of the quality of working, the shortening of working time and so on, but still many robots have their specialized functions. The expectation of present day society to robots has been diversified, and the technical development of high function robots is advanced positively. In this report, the recent examples of the high function robots developed for PWR power stations with the support of technical progress and the trend of the technical development are explained. The needs and seeds of maintenance robot development are discussed. As the examples of heightening the functions of maintenance robots, the next generation ultrasonic testing machine highly advanced by sensor technology and size and weight reduction mechanism technology, the intelligent monitoring system for welding using AI technology and other manpower-saving robots are shown. (K.I.)

  13. Economy aspect for nuclear desalination selection in Muria Peninsula using 1000 MWe PWR

    Full text: An assessment of economy aspect for nuclear desalination selection has been carried out. This study is done to explore any possibility to utilize co-generation concept of desalination, because there is a plan to introduce nuclear power plants (NPP) into Indonesia's electricity grid. A comprehensive study on different energy sources shows that NPP is economically and technically viable to be introduced into the grid in 2016/2017. The candidate site is Muria Peninsula in Central Java. Currently, the total of install electricity capacity is about 29.083 GWe, and it is estimated that electricity energy growth is about 7.1% per year. The install capacity in Java is about 23 GWe (65% of national capacity). With economic growth projection is about 6%, therefore in the 2025, it is needed electricity energy about 70 GWe, so electricity demand increase 2000 MWe per year. Therefore, a PWR of 1000 MWe will coupled with a desalination plant of MSF (Multi-stage Flash Distillation), MED (Multi-Effect Distillation) and RO (Reverse Osmosis). The costs of water production for the Multi Stage Flash Distillation (MSF), Multi Effect Distillation (MED) and Reverse Osmosis (RO) desalination process coupled to PWR 1000 MWe would be compared. The objectives of the economic evaluation is to help the decision-maker to eventually implement an integrated nuclear desalination plant, generating both electricity and fresh water. Economic analysis of water cost are performed using a computer program issued by the IAEA, DEEP-3.1. In this study, option for turbine scheme is set as extraction and back pressure. Options for specific carbon tax, thermal steam compression and backup heat are not used. Construction cost for NPP is assumed to be 2600 $/kW, production capacity 2.750 m3/d, interest rate 5%, construction cost for MSF 1200 $/m3/d, MED 900 $/m3/d and RO 700 $/m3/d, ratio of recovery RO 45%, top brine temperature for MED 65 deg. C and MSF 110 deg. C. The results of the performed case

  14. Krsko NPP Periodic Safety Review program

    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)

  15. Treatment of NPP wastes using vitrification

    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)

  16. Optimizing NPP performance and service life

    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

  17. Issues of risk management during NPP operation

    The paper outlines risk management issues during safety assessments of nuclear facilities and summarizes international experience in NPP risk management in different countries. The need is also considered to elaborate risk management and optimization procedures for Ukrainian NPPs

  18. Analysis of safety culture at Rovno NPP

    The main concepts of safety culture which relate to safety increase in reactor unit operation, their reliable work, high qualification of personnel and personal responsibility of operators are developed. They will be introduced at the Rovno NPP

  19. Development of a Computer Program for an Analysis of the Logistics and Transportation Costs of the PWR Spent Fuels in Korea

    It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU

  20. Training and qualification of NPP personnel in Slovenia. National summary

    Slovenia has two nuclear installations, one is experimental reactor TRIGA Mark 2, located at Jozef Stefan Institute near Ljubljana, and the second is Nuclear Power Plant Krsko, a two loop Westinghouse PWR. NPP Krsko was recently modernised which included installation of new steam generators and acquisition of a new full scope replica simulator. Slovenia is facing similar problems in nuclear field as other countries with nuclear programs. Major ones are establishment of open energy market, workforce ageing and long-term radioactive waste storage. Training in nuclear field was established at the very beginning of the nuclear program in Slovenia (which was at that time a part of former Yugoslavia). This was started in close cooperation between NPP Krsko and Jozef Stefan Institute (JSI) in late seventies. Training organisation and programs evolved over the two decades. During the years, in addition to the training of plant Personnel, Vienna (AT) the Nuclear Training Centre in Ljubljana became the leading institution for training of medical and industrial personnel in radiation protection, they have established very popular nuclear information centre and they are also internationally well known for their organisation of international workshops and courses as an IAEA Regional Resource Centre. Development of the training programs at the plant culminated with the introduction of full scope simulator and maintenance training centre in the year 2000. Two basic nuclear training programs are established for workers in nuclear field, which are conducted jointly by NPP Krsko and Nuclear Training Centre at JSI. One is an eight-week program called Basic Nuclear Technology Course and the second called Nuclear Technology Course. NPP Krsko full scope simulator is, in addition to standard simulation range, equipped with severe accident simulation capability. This has been achieved by integrating EPRI code MAAP 4 within the CAE (simulator vendor) models. MAAP 4 was selected as it is

  1. Geological evaluation of the Paks NPP site

    The geological evaluation of the site of nuclear power plant constitutes the basis for the assessment of seismic hazard important in terms of the NPP safety. Its re-evaluation is imperative because of the new safety requirements and the new scientific knowledge. Geological evaluation of the Paks NPP site is presented in this paper. Based on seismotectonic evaluation and the seismological data, the seismic hazard of the plant site was determined by using both probabilistic and deterministic methods

  2. Ignalina NPP Safety Analysis: Models and Results

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

  3. Training human resource for NPP in Vietnam

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

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

    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)

  5. Visaginas NPP Project Regional Approach: Lithuania

    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

  6. Suomi NPP Ground System Performance

    Grant, K. D.; Bergeron, C.

    2013-12-01

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

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

    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

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

    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)

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

    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

  10. Thermodynamic modelling of PWR coolant

    Corrosion products released from PWR and VVER primary circuit surface oxides are transported in the coolant to the core, where they deposit and are activated to form radioactive corrosion products, which can be re-released to re-deposit on out-of-core surfaces. Spinel solubilities vary with the pH, temperature and sometimes the hydrogen concentration of the coolant. This paper describes the development of an equilibrium thermodynamic model to predict such changes, and discusses the extent of the available solubility data for Fe, Ni, Co and Zn oxides. Results are described on the relative solubility of Fe and Ni under both normal operating conditions and during shutdown/start-up, and on the relative stabilities of stoichiometric and non-stoichiometric zinc ferrite spinels. Comparison of the calculated corrosion product concentrations with reactor measurements indicates that, in reactors with low Ni content in the steam generator alloys, the concentration of Ni in the coolant is limited by its availability in the surface oxide. In reactors with high-Ni alloys, the circulating Ni concentrations may be dominated by colloidal material. The calculated changes in Ni and Fe concentrations during the acid-reducing phase of shutdown are in reasonable agreement with measurements from Sizewell B. The paper highlights the need for a more comprehensive open corrosion product data base, the need to consider both boiling and radiolysis in the core on corrosion product solubility in different parts of the primary circuit and, finally, the importance of kinetic factors at low temperature behaviour during shutdown and start-up. (author)

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

    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

  12. Selection of compositions for the cementation of liquid radioactive waste of Kudankulam NPP and Volgodonskaya NPP

    The purpose of this work is the selection of formulations for the cementation of liquid radioactive waste of Kudankulam NPP and Volgodonskaya NPP. The simulators of the following radioactive waste have been used for the works: concentrated still bottoms (CSB) with saline content 600-800 g/l, sludge, pulps of ion-exchange resins (IER), activated carbon, titanium and ion-exchange sorbents of Kudankulam NPP and concentrated still bottoms with saline content 900 g/l, pulps of ion-exchange resins, sludge of Volgodonskaya NPP. For Kudankulam NPP there was made a separate research of the cementation of each type of waste and also joint cementation of concentrated still bottoms and ion-exchange resin. For Volgodonskaya NPP - joint cementation of CSB and IER or sludge. The properties of the compounds were determined, which are regulated by GOST R 51883-2002, spread ability and setting time of the cement grouts. The study has shown that as a main component of the combined binding material for the cementation of low level radioactive waste (LRW) of Kudankulam NPP and Volgodonskaya NPP, the usage of Portland cement is preferable. As additives for the binding materials it is better to use lime and bentonite clay powder. Maximal inclusion of LRW into the compound when using these materials will be (% of the compound weight): CSB: 30%, sludge - 14%, IER - 14%, activated carbon - 18%, titanium sorbent - 20%, ion-selective sorbent - 14%

  13. Upgrade of Common Cause Failure Modelling of NPP Krsko PSA

    Over the last thirty years the probabilistic safety assessments (PSA) have been increasingly applied in technical engineering practice. Various failure modes of system of concern are mathematically and explicitly modelled by means of fault tree structure. Statistical independence of basic events from which the fault tree is built is not acceptable for an event category referred to as common cause failures (CCF). Based on overview of current international status of modelling of common cause failures in PSA several steps were made related to primary technical basis for methodology and data used for CCF model upgrade project in NPP Krsko (NEK) PSA. As a primary technical basis for methodological aspects of CCF modelling in Krsko PSA the following documents were considered: NUREG/CR-5485, NUREG/CR-4780, and Westinghouse Owners Group documents (WOG) WCAP-15674 and WCAP-15167. Use of these documents is supported by the most relevant guidelines and standards in the field, such as ASME PRA Standard and NRC Regulatory Guide 1.200. WCAP documents are in compliance with NUREG/CR-5485 and NUREG/CR-4780. Additionally, they provide WOG perspective on CCF modelling, which is important to consider since NEK follows WOG practice in resolving many generic and regulatory issues. It is, therefore, desirable that NEK CCF methodology and modelling is in general accordance with recommended WOG approaches. As a primary basis for CCF data needed to estimate CCF model parameters and their uncertainty, the main used documents were: NUREG/CR-5497, NUREG/CR-6268, WCAP-15167, and WCAP-16187. Use of NUREG/CR-5497 and NUREG/CR-6268 as a source of data for CCF parameter estimating is supported by the most relevant industry and regulatory PSA guides and standards currently existing in the field, including WOG. However, the WCAP document WCAP-16187 has provided a basis for CCF parameter values specific to Westinghouse PWR plants. Many of events from NRC / INEEL database were re-classified in WCAP

  14. Operation Aspect of the Main Control Room of NPP

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

  15. Decommissioning study of Forsmark NPP

    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

  16. Decommissioning Study of Oskarshamn NPP

    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

  17. Decommissioning study of Forsmark NPP

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

    2013-06-15

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

  18. Decommissioning Study of Oskarshamn NPP

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

    2013-06-15

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

  19. Activity transport models for PWR primary circuits

    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.)

  20. Program of monitoring PWR fuel in Spain

    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)

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

    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.

  2. Passive safety systems for the next generation of NPP's main R and D activities

    Containment Cooling and Depressurization of the Reactor Coolant, two major topics of mitigation of consequences of beyond design basis core damage accidents are dealt with by passive systems co-developed by Ansaldo and ENEL for the next generation of NPP's in the frame of international co-operation. A Passive Containment Cooling System (PCCS) concept consisting of modular loops, each with inner heat exchanger, outer condenser and interconnecting piping, has been developed for application to PWR units with dual concrete (EUR requirement) containment type. Two versions of the inner heat exchanger have been designed; the first one, under development by ENEL, features a compact tube-bundle with top-bottom natural draught of the air-steam mixture; the second one, under development by Ansaldo, consists of water-jacket modules embedded in the concrete containment. The key-components, the Isolation Condenser and the Passive Containment Cooler, of two passive systems for application to the SBWR, the advanced BWR of GE, for the control of respectively reactor and containment pressure have been developed, designed and tested on full-scale prototypical units. Depressurization of the Reactor Coolant by injection of cold borated water into the steam plenum is the result of the Passive Injection and Depressurization System (PIDS), a completely passive concept, applicable to both PWR and BWR designs

  3. Potential issues related to emergency core cooling system strainers performance at boiling water reactors: Application to Cofrentes NPP (Spain)

    Rubio, Rafael, E-mail: rrubio@iberdrola.es [Iberdrola Generación Nuclear S.A., Madrid (Spain); Jimenez, Gonzalo [Universidad Politécnica de Madrid (Spain)

    2014-08-15

    Highlights: • Design of the ECCS strainers introduced a reasonable margin for operation in BWRs. • Studies are addressing the effects of post-LOCA debris on ECCS in Cofrentes NPP. • The head loss due is at most half of the limited head loss for the ECCS strainer. • The NPSH required is at least three times lower than the NPSH available. - Abstract: From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available.

  4. Potential issues related to emergency core cooling system strainers performance at boiling water reactors: Application to Cofrentes NPP (Spain)

    Highlights: • Design of the ECCS strainers introduced a reasonable margin for operation in BWRs. • Studies are addressing the effects of post-LOCA debris on ECCS in Cofrentes NPP. • The head loss due is at most half of the limited head loss for the ECCS strainer. • The NPSH required is at least three times lower than the NPSH available. - Abstract: From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available

  5. Economic aspects of the Belene NPP project continuation - analysis of international data

    The aim of this study is to estimate the probable cost of electricity which would be produced in the projected Belene NPP in Bulgaria. The construction of the Belene NPP started in 1987 and were stopped in 1990 due to financial and socio-ecological reasons. The analysis is based on comparative data on electricity production in PWR reactors Stendal, Temelin, Zaporozhe-6, Rovno-4 and Khmelnitskij-2,3,4. The capital costs, costs for operation and repair, costs for fuel/waste management and decommissioning costs are taken into account. It is stressed that the capital costs are highly increased due to elevated requirement for safety and reliability, equipment aging and spent fuel/ waste management. In most countries the electricity cost value is in the range 0.025 - 0.105 US$/kWh. It is concluded that the cost of electricity produced in Belene-1 reactor would be about 0.040 US$/kWh plus 0.005 - 0.010 US$/kWh for spent fuel/waste expenditures. The author's opinion is that the completion of the plant would be economically unprofitable

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

    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)

  7. Applying a small NPP in the Argentine mining industry

    The CAREM 25 reactor project is a small PWR nuclear power plant of 27 MWe, based on advanced concepts: a self-pressurized integral primary with natural convection of the coolant and a more simple and reliable general design. The CAREM concept has many advantages as a power generator in small electrical grids. Besides, there are some non-electrical applications under consideration, since a co-generation scheme seems very interesting from the economical point of view. In this category two alternatives have been considered: a standard desalination facility and a process plant in the mining industry. In this paper, a conceptual analysis of the second alternative is presented. Mining is a branch of the domestic industry that has shown a remarkable growth in the past three years mainly due to a steady inflow of foreign investments (about two billion dollars for that period). And one of the most attractive markets is in the extraction and manufacturing of non-ferrous minerals, coming from deposits in the northwest of Argentina: sodium sulfate, lithium salts, and boron compounds. Nevertheless it faces an unsolved problem in the energy high prices due to the fact that the production sites are located in remote areas where the only achievable energy source is the transportation of fuel oil. In this scenario, a small NPP may be a competitive source of process heat and electricity, with enough autonomy to uncouple fuel requirements from production strategies. The present study analyses the possible application of the CAREM concept in the non-ferrous mining industry of the Northwest of Argentina, considering a co-generation scheme. The main results of this analysis and the inherent advantages of the approach, show that the alternative may be feasible both from the technical and the economical points of view. (author)

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

    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.

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

    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

  10. Integrated ageing management of Atucha NPP

    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)

  11. Environmental impact of the NPP Krsko

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

  12. Integrated ageing management of Atucha NPP

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

    2013-07-01

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

  13. Regulatory aspects of NPP safety

    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)

  14. Kozloduy NPP intranet portal, Bulgaria

    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

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

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

    2008-01-01

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

  16. Simulation model of a PWR power plant

    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)

  17. PWR reactors for BBR nuclear power plants

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL)

  18. Full MOX core design for PWR

    Full MOX core design for APWR was analyzed in nuclear design, fuel integrity analysis, thermal hydraulic design and safety analysis et. al. Feasibility of Full MOX core was confirmed from these analyses without any large modifications. Full MOX PWR core has very good characteristics in which single Pu content in an assembly, burnable poison free, higher burnup and longer cycle operation are feasible. (author)

  19. An evaluation of tight - pitch PWR cores

    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. Some problems of NPP construction base improvement

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

  1. Safeguards at Kozloduy NPP - Experience and expectations

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

  2. Safety upgrading program in NPP Mochovce

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

  3. Knowledge management during decommissioning of Chornobyl NPP

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

  4. Medical consequences of NPP and TPP operation

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

  5. Radioactive waste problems in the Kozloduy NPP

    An average volume of 1400 m3 a year of solid radioactive waste (RAW) is generated in the Kozloduy NPP. The adopted waste processing sequence is collection, sorting and compaction with a 1000 tons force providing decrease in volume by factor of 15. A temporary storage facility at the Kozloduy NPP is licensed by ISUAE and CPPUAE. The treatment of liquid wastes is performed by Westinghouse formula and a technology using an automated solidification system. Contaminated oils are burned using an oil incinerator. A special 2-year programme for RAW management is being developed

  6. Lifetime evaluation of Bohunice NPP components

    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)

  7. Ignalina NPP pre-decommissioning projects

    Description of the main projects for the preparation to the decommissioning of unit 1 of Ignalina NPP is presented. These projects are to be financed by international donors as one of the conditions to shutdown unit before the year 2005. These projects were presented during Donors conference held in 21-22 June 2000 in Vilnius. The conference was organized jointly by Lithuanian Government and European Commission. Projects are devoted to the construction of radioactive waste management facilities and improvement of existing waste management practices at Ignalina NPP as well for the general management of decommissioning process preparation of necessary documentation

  8. REWET, PWR LOCA accident experiments

    1 - Description of test facility: The REWET-II facility was designed for the investigation of the reflooding phase of a LOCA. The main design principle is the accurate simulation of the rod bundle geometry and the primary system elevations. This is necessary in order to have the correct flow channels and hydrostatic pressures for the reflooding process. The reactor vessel is simulated by a stainless steel U-tube construction consisting of downcomer, lower plenum, core and upper plenum. The primary loops contain a pipe simulating the broken loop and a connection line between the upper plenum and the downcomer simulating five intact loops. The containment is simulated by a pressure vessel (not in scale). Steam generators and primary pumps are simulated with flow resistances. The ECC-water can be injected to the downcomer and/or to the upper plenum by a pump or from an accumulator. All the elevation in the reactor vessel simulator are scaled to 1:1 (except the reactor upper head). The scale of the volumes and flow areas is 1:2333 referring to the number of the fuel rod simulators in the facility and the fuel rods in the reference reactor. The rod bundle is either in a hexagonal shroud, the inside distance of the opposite walls is 54.3 mm and the wall thickness 2 mm, or in a round shroud, the inner diameter is 66.0 mm and the wall thickness 2 mm. The simulation of the fuel-rod bundle consists of 19 indirectly electrically heated simulator rods. The heating coils are inside stainless steel claddings in magnesium oxide insulation. The heated length, the outer diameter and the lattice pitch of the fuel-rod simulators as well as the number (= 10) and construction of the rod bundle spacers are the same as in the reference reactor. The upper ends of the rods are attached to the upper tie plate. 2 - Description of test: Pressurized water reactors in use in Finland reactors have certain unique features which make them different from most other PWR designs. The 6 horizontal

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

    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

  10. Hitachi DCS emulator design to support NPP simulator implementation

    Nuclear Power Plant (NPP) simulators are the main means for operator training and as such are a crucial part of the NPP operation life-cycle.Efficient development and testing of NPP software and design concepts require a robust platform mirroring the design and configuration of the operating plant.DCS Emulator and full-scope simulator (FSS) technologies support both these objectives for the entire NPP life-cycle allowing users and operators to implement, test and use actual control and information software applications and designs. This paper describes Hitachi's latest simulator development and challenges in implementing a DCS emulator to provide a code emulation platform for developing NPP software. (author)

  11. Principles of tariff determination for NPP electric power generation

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

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

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

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

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

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

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

  15. Radiation monitoring of the environment - Kozloduy NPP

    The questions concerning environmental protection are always concomitant to the electrical industry of a country. Besides, social opinion is very susceptible to the problems of nuclear energy production, especially in relation to the qualitatively new pollutants and nuclear production hazards. What is pointed out is the place of the nuclear power station in the energy production system and its contribution to the restriction of gaseous releases as one of the global ecological problems. The paper presents the programme for radiation monitoring of the environment and the results of its realization, as they characterize the environmental impact of NPP Kozloduy for the period of its operation. An overview of the current environmental state in the vicinity of NPP Kozloduy is made. It includes the results concerning radioactivity in major environmental components (air, surface waters, soils, milk, fish), as well as gamma dose rate equivalent around NPP. A comparison between the 2000's results and other for long-term studies is presented. An evaluation of the population exposure as results of NPP Kozloduy operation is made. The absence of statistically reliable changes in the radiation characteristics of the environment and the steady process of safety improvement are a real prerequisite for the development of nuclear energy production in our country as an ecologically consistent activity, based on socially acceptable risk. (authors)

  16. Safety culture development at Daya Bay NPP

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

  17. Safety upgrading of Bohunice V1 NPP

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

  18. Development of NPP safety regulation in Russia

    The presentation describes the organisation scheme of Russian safety regulatory bodies, their tasks and responsibilities. Legislative and regulatory basis of NPP safety regulations rely on the federal laws: Law on the Use of Nuclear Energy and Law on Radiation Safety of the Population. Role of international cooperation and Improvement of regulatory activities in Russia are emphasised

  19. Thermohydraulic calculation of WWER-type NPP

    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

  20. Intelligent alarm-processing system for NPP

    Information on developing the intelligent alarm-processing system for NPPs with BWR reactors, which makes it possible to reduce the information load for the operators through the information volume optimization, related to identification of failures in the NPP operation, is presented. Calculational principles and methodological constituents for processing alarm signals are considered. Description of the system and simulation check results are presented

  1. NPP Krsko fire protection action plan

    This paper describes the Fire Protection Action Plan which prioritized proposed fire protection modifications from recommendations reported in the NPP Krsko Fire Hazards Analysis - Safe-Shutdown Separation Analysis (SSSA), the ICISA Analysis of Core Damage Frequency Due to Fire at the Krsko Nuclear Power Plant, and the Operational Safety Review Team (OSART) reports using a risk-based cost/benefit methodology. (author)

  2. Works on ageing management for Cernavoda NPP

    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)

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

    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)

  4. RHR system reliability analysis of Krsko NPP

    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)

  5. Cernavoda NPP training programs The paper presents a general assessment of Cernavoda NPP personnel training programs,

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

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

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

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

  7. Sitting Safety Aspects of Second Romanian NPP

    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

  8. Progress of PWR reactor fuels: OSIRIS equipments

    The experimental reactor Osiris situated at the Saclay Nuclear Centre is a reactor fitted with tests and monitoring facilities. Of the pool and open core type, it can test the test fuel of PWR power stations under high neutron flux. The characteristic stresses of the operating states of power reactors can be reproduced in experimental devices suited to the various study subjects, be this the creep and deformation of zircaloy claddings, the behavior of fuel rods to power ramps, to load following, to remote regulation, to the cooling state in double phase or just analytical tests. The experimental irradiation devices extend from the single static coolant capsule, such as the NaK alloy, to the dynamic coolant test loop that operates in the cooling conditions representative of PWR's including water chemistry. Ancillary devices make it possible to carry out examinations and non-destructive testing: immersed neutron radiography, gamma scanning visualization monitoring device, eddy currents, profilometering

  9. Horizontal Drop of 21- PWR Waste Package

    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in-terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 1 1) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design

  10. Thorium fuel cycle study for PWR applications

    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.