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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. 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

  13. 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

  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. 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)

  9. 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)

  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. 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.)

  12. 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)

  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 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

  16. 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.)

  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. 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

  6. 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)

  7. 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

  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. 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)

  11. 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)

  12. 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).

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

    Thunem, Harald P-J; Zhang, Xinxin

    2015-01-01

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

  14. 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)

  15. 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)

  16. 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)

  17. 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

  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. 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

  16. 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

  17. 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.

  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. 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

  3. 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

  4. 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

  5. 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

  6. 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

  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. 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)

  13. 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

  14. 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)

  15. 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)

  16. 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)

  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. 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)

  19. 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

  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. 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)

  3. 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.

  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. 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

  8. 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)

  9. 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

  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. 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

  15. 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)

  16. 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)

  17. 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

  18. 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

  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. 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)

  3. 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)

  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 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

  8. 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

  9. 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)

  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. 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.)

  16. 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)

  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. 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

  20. 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

  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. 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)

  6. 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

  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. 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)

  11. 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)

  12. 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)

  13. 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

  14. 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)

  15. 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)

  16. 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)

  17. 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)

  18. 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.)

  19. 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.

  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. 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

  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. 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

  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. 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 ...

  11. 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 ...

  12. 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

  13. 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)

  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. 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

  16. 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

  17. 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)

  18. 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)

  19. 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

  20. 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

  1. 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

  2. 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)

  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. 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)

  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. 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

  9. 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

  10. Thorium fuel cycle study for PWR applications

    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 UO2 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 ThO2-PuO2 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 UO2 fuel. (author). 6 refs., 3 tabs., 6 figs

  11. 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. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only. 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. 11) 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

  12. 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.

  13. Cernavoda NPP simulator - next generation

    Demand for extending the amount of training and scope for Cernavoda Unit 1 as well as the new trend in the simulator owners world, led to a change in the Romanian philosophy of simulator specification. Up to now the training was conducted on a Full Scope simulator, a 1:1 replica of Cernavoda Unit 1 reference plant. The present task is to define the simulation facilities and structure capable to meet the requirements for training, qualification and licensing of personnel for both Cernavoda Unit 1 and Unit 2. Obviously, the Cernavoda Unit 2 belongs to the same technological family but has rather different control room layout. Since this target requires a new simulator the costs would be rather high in accordance to the degree of automation of Cernavoda NPP. Therefore, depending on training requirements and financing, the Cernavoda Unit 1 simulator modernization, which also provides an alternative to full scope control room simulator, may be a viable option. Therefore the solution that with discuss for Cernavoda training extension is the migration of Cernavoda Unit 1 simulator to state-of-the-art. Consequently, the Cernavoda Unit 1 simulator modernization task will be organized as project including the following major items: 1. Rehost existing U1 simulation software from VAX 4500 to: - Best commercial multi-processor server for simulation server (HP, O/S Linux); - Best commercial single processor PC for I/O communications (HP, O/S Linux); 2. Replace DCC with enhanced emulated version: Best commercial individual PC for DCC emulation (HP, O/S Windows); Support for actual keyboards; Replacement of RAMTEK System and CONRAC Monitors with X terminals or PC's; 3. Conversion of AutoCAD-based panel graphic pages to RAVE-based; 4. Install the required software tools for developing enhanced simulation modules; 5. Replace the simulation modules with advanced modules; 6. Replace the present Windows Instructor Facilities with ISIS; 7. Development of a selection of MCR-U1 virtual

  14. Safety Analysis Report for Ignalina NPP

    In December 1994 an agreement was signed between the European Bank for Reconstruction and Development and the Republic of Lithuania for the grant of 32.86 MECU for the safety Improvement at Ignalina NPP. One of the conditions for the provision of the grant, was a requirement for an in-depth analysis of the safety level at Ignalina NPP in the scope and according to the standards acceptable for a western nuclear power plant, and to publish a Safety Analysis Report (SAR). The report should investigate and analyze any factor that could limit a safe operation of the plant, and provide recommendations for actual safety improvements. According to the agreement, Lithuania had to finalize the SAR until 31 December, 1995. The bank has also organized and financed investigation of safety at Ignalina NPP and preparation of the SAR. EBRD made an agreement with Sweden's Vattenfall, which subcontracted well-known companies from Canada, USA, Germany, etc., and also the Russian Research and Development Institute of Power Engineering (NIKIET), reactor designer of Ignalina NPP. The SAR is a very comprehensive document and contains about 8000 pages of text, diagrams and tables. The main findings of the SAR are provided in the article. A large number of discrepancies with modern rules and western practices was detected, but they were not proved to be serious enough to require reactors shutdown. Based on the recommendations of the SAR Ignalina NPP has worked out Safety Improvement Program No. 2 (SIP-2), which is planned for three years and will cost 486 MLT. (author)

  15. Space-dependent dynamics of PWR

    The azimuthal dependent reactor dynamics coupled to thermohydraulics are studied by using the neutron-flux and coolant temperature signals measured at an actual PWR. The second azimuthal mode of neutron-flux fluctuation was found, and the coupling of the mode to thermohydraulics of the coolant was suggested. The coherent coolant flow in the reactor core seems to sustain this spatial oscillation mode. (authors)

  16. Sensitivity analysis of a PWR pressurizer

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  17. PWR fuel behavior: lessons learned from LOFT

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

  18. Optimum fuel use in PWR reactors

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.)

  19. Chemical and radiochemical specifications - PWR power plants

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

  20. Steam shut-off valves for PWR type reactors

    Fast acting closure means are requested in PWR type reactors as well as in BWR to safely shut-off the live steam at the turbine input in the event of accident. The design and control system of steam shut-off valves acted by the fluid system and intended for PWR type reactors, are described. The role of these valves in a PWR is discussed with the specified requirements involved

  1. Modelling activity transport behavior in PWR plant

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

  2. Nondestructive examination requirements for PWR vessel internals

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  3. Dismantling and decommissioning experience of commercial PWR

    Regarding the relatively youthness of FRAMATOME PWR's in operation none of these reactor needs to be decommissioned before 1992. However feasibility studies have been carried out by FRAMATOME for an on site entombment of active components and heavy equipments. In the past, partial dismantling of the reactor internals of the CHOOZ reactor: PWR of 320 MWe and a complete removal of the thermal shield protecting the reactor vessel were conducted successfully. After repair, the reactor power output has been upgraded of 10% and the reactor operates satisfactorily since 1970. More recently the discovery of scarce defects affecting centering pins of control guide tube located in the upper reactor internals of 900 MWe plants has initiated the construction of several ''Hot stand equipments'' for the systematic replacement of these centering pins. FRAMATOME is presently actively studying possible options consisting either to extend the plant life beyond its initial licence life, or to convert classical PWR into an advanced reactor more economical in terms of uranium consumption

  4. Pu-breeding feasibility in PWR

    This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO2 or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived

  5. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  6. A historical survey of the Ignalina NPP

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

  7. Analyses and estimation of insulation material release 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 und maintain strainer clogging effects after a loss-of-coolant accident for all reactor types The strainer performance must be adequate LOCA and post-LOCA. Therefore clogging has to be avoided basically or at least minimized to acceptable pressure loss rates with respect to the needed suction head of the emergency cooling pumps (NPSH) and to the designed differential pressure of the strainers themselves. These aspects might become safety relevant for the emergency core cooling procedure after LOCA if not acceptable conditions are expected within the accident control phase or later during the post accident phase. In any case counter actions, technical provisions and procedures have to be planned and described in the operational manuals. Especially for the German Pressurized Water Reactors (PWR) a program was launched by the German Utilities together with the plant manufacturer in order to carry out theoretical and experimental analyses about the behavior of released insulation particles of mineral wool and/or other materials such as micro porous materials. The program was run over years and technically finalized in 2008. Main aspects and results cover all effects of the handled materials, of their specific behavior and of the interactions with the strainers as part of the emergency core cooling systems. The intensive investigations covered all aspects to be considered for an overall understanding and analyses of a loss of coolant event (LOCA) with an induced release of insulation particles and accompanying effects which may affect the strainer function regarding to clogging and the corresponding counteractions. Therefore starting with the

  8. Revision of Krsko NPP Quality Assurance Plan

    International standards from nuclear power plant operation area are being frequently upgraded and revised in accordance with the continuous improvement philosophy. This philosophy applies also to the area of Quality Assurance, which has also undergone significant improvement since the early 1950s. Besides just nuclear industry, there are also other international quality standards that are being continuously developed and revised, bringing needs for upgrades also in the nuclear application. Since the beginning of Krsko NPP construction, the overall Quality Assurance program and its applicable procedures were in place to assure that all planned and systematic actions necessary to provide adequate confidence that an item or service will satisfy given requirements to quality, are in place. The overall requirements for quality as one of the major objectives for Krsko NPP operation are also set forth in the Updated Safety Analyses Report, the document that serves as a base for operating license. During more than 30 years of Krsko NPP operation, the quality requirements and related documents were revised and upgraded in several attempts. The latest revision 6 of QD-1, Quality Assurance Plan was issued during the year 2011. The bases for the revision were: Changes of the Slovenian regulatory requirements (ZVISJV, JV5, JV9?), Changes of Krsko NPP licensing documents (USAR section 13?), SNSA inspection requirements, Changes of international standards (IAEA, ISO?), Conclusions of first PSR, Implementation of ISO standards in Krsko NPP (ISO14001, ISO17025), Changes of plant procedures, etc. One of the most obvious changes was the enlargement of the QA Plan scope to cover interdisciplinary areas defined in the plant management program MD-1, such as Safety culture, Self-assessment, Human performance, Industrial Safety etc. The attachment of the QA Plan defining relationships between certain standards was also updated to provide matrix for better correlation of requirements of

  9. Optimizing NPP performance and service life

    Generally, at this time Hungary has not any delayed nuclear projects or tasks having special importance to need urgent measures or significant technical investments. Nevertheless, the power reactors produce significant amount of spent fuel and radioactive waste. In Hungary there is only one facility for radioactive high level waste management near Budapest. The low- and intermediate level radioactive waste has been stored on site of PAKS NPP from the start of operation in safe conditions. In 1993, Hungary launched a national programme to solve the problems of radioactive waste management. A new, potential site was identified in granite host rock for a below-surface repository. Ongoing investigations at this site are being carried out; a four-year research programme is in progress. It is remarkable, that at the beginning the first investigation programme met some public resistance some years ago causing a short period delay. This negative attitude has been changed because the NPP offered working places and infrastructure development for the near situated village inhabitants. In PAKS NPP all the modifications, subsystem modernisation, replacements, special maintenance activities, etc. are time-scheduled. At present, the average time of the yearly refuelling and general maintenance is about 30 days in PAKS NPP. From economical point of view, the short period of the outage is desirable. Nevertheless, the first goal must be in the maximum level: the nuclear safety. To reach the optimal balance between the two requirements, NPP PAKS introduced some methods, organisational systems, and maintenance procedures. The maintenance is planned with computer, which determines even the smallest steps in optimal sequences. It has been introduced some years ago that some types of maintenance can be realised in time of operation before the outage of the reactor. The timetable of the outages is the object of regulatory approval. In cases of certain important delays, the NPP should

  10. NPP bulk equipment dismantling problems and experience

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

  11. NPP Temelin. Status of safety improvements

    The WWER-1000 Temelin NPP under construction has been subjected as other NPPs of the same type to numerous project reviews resulting in quite a number of recommendations for design changes. Results of the IAEA mission to review the resolution of WWER-1000 safety issues at Temelin NPP are cited in this paper. The main conclusions emphasize that a combination of eastern and western technology and practices led to safety improvements in comparison with the international practices. Plant managers are clearly committed to implementation of operational programs which are consistent with effective western operational safety practices. Considerable effort remains to bring planned programs to successful implementation, in particular in meeting the need to foster strong safety culture among all personnel

  12. Review of site data for Cernavoda NPP

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

  13. SAT for NPP personnel training in USA

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

  14. LTO License Application Project NPP Borssele

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

  15. Intelligent system for accident identification in NPP

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

  16. Design safety improvements of Kozloduy NPP

    Design safety improvements of Kozloduy NPP, discussed in detail, are concerned with: primary circuit integrity; reactor pressure vessel integrity; primary coolant piping integrity; primary coolant overpressure protection; leak before break status; design basis accidents and transients; severe accident analysis; improvements of safety and support systems; containment/confinement leak tightness and strength; seismic safety improvements; WWER-1000 control rod insertion; upgrading and modernization of Units 5 and 6; Year 2000 problem

  17. Valves for TPP and NPP in France

    Valves produced by the leading French firms for TPP and NPP are described in brief on the base of prospects presented at the ''Oil and Gas-83'' exhibition in Moscow. The method of testing for leak-tightness applied in the production of the valves is considered as wetl as the main parameters of working fluids for which some types of the valves are intended

  18. Design floor response spectra for Kozloduy NPP

    This presentation contains a detailed review of the design response spectrum for the floor of Kozloduy NPP Units 5 and 6. Design model was developed according to the plan to build this type od nuclear power plants in different regions. Calculations were done with a set of accelerograms, which includes artificial and already known recordings of earthquakes. Envelope response spectrum of the set of accelerograms is presented

  19. Dukovany NPP fuel cycle benchmark definition

    The new benchmark based on Dukovany NPP Unit-2 history is defined. The main goal of this benchmark is to compare results obtained by different codes used for neutron-physics calculation in organisations which are interested in this task. All needed are described in this paper or there are given references, where it is possible to obtain this information. Input data are presented in tables, requested output data format for automatic processing is described (Authors)

  20. Supercompaction of radioactive waste at NPP Krsko

    The problem of radioactive waste management is both scientifically and technically complex and also deeply emotional issue. In the last twenty years the first two aspects have been mostly resolved up to the point of safe implementation. In the Republic of Slovenia, certain fundamentalist approaches in politics and the use of radioactive waste problem as a political tool, brought the final radioactive repository siting effort to a stop. Although small amounts of radioactive waste are produced in research institutes, hospitals and industry, major source of radioactive waste in Slovenia is the Nuclear Power Plant Krsko. When Krsko NPP was originally built, plans were made to construct a permanent radioactive waste disposal facility. This facility was supposed to be available to receive waste from the plant long before the on site storage facility was full. However, the permanent disposal facility is not yet available, and it became necessary to retain the wastes produced at the plant in the on-site storage facility for an extended period of time. Temporary radioactive storage capacity at the plant site has limited capacity and having no other options available NPP Krsko is undertaking major efforts to reduce waste volume generated to allow normal operation. This article describes the Radioactive Waste Compaction Campaign performed from November, 1994 through November, 1995 at Krsko NPP, to enhance the efficiency and safety of storage of radioactive waste. The campaign involved the retrieval, segmented gamma-spectrum measurement, dose rate measurement, compaction, re-packaging, and systematic storage of radioactive wastes which had been stored in the NPP radioactive waste storage building since plant commissioning. (author)

  1. NPP EQUIPMENT LIFETIME MANAGEMENT UNDER AGEING

    Olga M. Gulina

    2012-01-01

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

  2. Perspectives of Living PSA in NPP Krsko

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

  3. Zoning Concept of The Muria NPP Area

    NPP development planning at Ujung Lemahabang site, Balong Village, Kemang District, Jepara Regency should be integrated with the regional development planning of Jepara Regency and Jawa Tengah Province due to the site is belong to the regency area. Regional Spatial Development Planning (RTRW) of Jepara is a master plan of Jepara regional development planning and it will be as a reference for all development implementation in Jepara Regency, including the NPP planning. In the NPP Development planning, public and environmental safety aspect are the major consideration that they should be accommodated. Both of normal operation and postulated accident case should ensure public safety against ionizing radiation hazard. If there will be an emergency situation, the spatial concept should ensure public safety considering population factor, geography, and spatial condition that will be able to make emergency countermeasure after the accident such as sheltering, evacuation, iodine stable distribution, decontamination, food bans, and relocation. Zoning is the appropriate concept to give safety assurance. The method to be used in the research is safety analysis using spatial approach. Exclusion zone will be recommended inside the ring of 1 km from nuclear reactor, while the low population zone will be recommended at 3.5 km radius from nuclear reactor. (author)

  4. Decommissioning database of V1 NPP

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

  5. Knowledge management in the NPP domain

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

  6. Tritium liquid effluents from the Krsko NPP

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

  7. Financial and organizational models of NPP construction projects

    The recent evolution of financial and organizational models of NPP projects can be truly reputed to open a new page of the world market of NPP construction. The definition of the concrete model is based mostly on specific cooperation backgrounds and current terms and conditions under which the particular NPP project is being evolved. In this article the most commonly known strategies and schemes of financing structuring for export NPP construction projects are scrutinized. Special attention is paid to the analysis of BOO/BOT models which are based on the public-private partnership. Most BOO/BOT projects in the power sector has Power Purchase Agreements (PPA) as an integral part of them. The PPA key principles are studied here as well. The flexibility and adaptability of the public-private partnership models for financing and organization of the NPP projects contributes substantially to the competitiveness of the NPP projects especially under current economic conditions. (orig.)

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

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

  9. Bilibino NPP: Operation experience and design lifetime extension

    The Bilibino NPP (BiNPP) has been operated since early 1974 near the town of Bilibino in Chukotka. A high effectiveness of the nuclear energy sources use under the rigorous conditions of the Russian Federation's North- East region has been demonstrated. BiNPP was designed as nuclear co-generation plant. Some specific features of the area where the BiNPP is sited have necessitated several original engineering solutions in the reactor plant development and in the design of the NPP. Their correctness has been confirmed by the operating experience. BiNPP consists of four power units of the same type. The BiNPP's installed capacity amounts to 48 MW, with simultaneous heat production of 78 MW. In the period when the Russian economy was stable (up to 1991), the plant capacity factor amounted to 85%, with that of operating availability of 90-92%. The BiNPP's economic parameters were considerably superior vs. those of local organic fuel fired power sources. This advantage appears to increase to date due to significant rise of prices for organic fuel brought to the area. The analysis of accidents with normal operation of safety systems and with their failures has revealed features of rather high inherent self-protection of the reactor. The reactors' failure-free operation ensures the high reliability of BiNPP as power source under extreme conditions of the Far North-East. The BiNPP 1st unit's design lifetime of 30 years is to end in January of 2004, and that of the 4th unit - in December 2006. The question of extending the operation lifetime of Bilibino NPP beyond the design limit was raised because the project of the BiNPP-2 (second construction stage) developed in 1992 proved not to be feasible due to the high construction cost. (author)

  10. Expression of NPP1 is regulated during atheromatous plaque calcification

    Nitschke, Yvonne; Hartmann, Simone; Torsello, Giovanni; Horstmann, Rüdiger; Seifarth, Harald; Weissen-Plenz, Gabriele; Rutsch, Frank

    2009-01-01

    Abstract Mutations of the ENPP1 gene encoding ecto-nucleotide pyrophosphatase/phosphodiesterase 1 (NPP1) are associated with medial calcification in infancy. While the inhibitory role of matrix proteins such as osteopontin (OPN) with respect to atherosclerotic plaque calcification has been established, the role of NPP1 in plaque calcification is not known. We assessed the degree of plaque calcification (computed tomography), NPP1 and OPN localization (immunohistochemistry) and expression (RT-...

  11. Present situation and development trend of NPP emergency operating procedure

    The present situation of the NPP emergency operating procedures and their inadequacy and the international development trend are briefly described as following: supplement or substitute event oriented emergency operating procedures by symptom oriented emergency operating procedures; using modern PC data management technology for production, modification, printing and display of NPP operating procedures; developing a artificial-intelligent NPP operating expert system which will help operator to govern accident with few human factor mistakes

  12. IAEA and problems of security. Specific principles of NPP safety

    Conceptual heart for specific principles of safety and its connection with fundamental ideas of the protection in depth and culture of safety are treated. Presented scheme involves such positions as choice of site for NPP, projecting of NPP, manufacture of equipment and building of station, commissioning into service, operation NPP, accident control, removal from service, readiness to accident. The presentation is reviewed of peculiar principles of safety in the IAEA INSAG-12 report: Basic principles of safety

  13. PWR Core 2 Project accident analysis

    The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16,500 kilograms of depleted natural uranium (as UO2), 139 kilograms of plutonium, 2.8 megacuries of mixed fission products, and 14 kilograms of Zircaloy-4 (cladding and hardware) into the 221-T Canyon Building. Event sequences for potential accidents that were considered included (1) leaking fuel assemblies, (2) fire and explosion, (3) loss of coolant or cooling capability, (4) dropped and/or damaged fuel assemblies, and extrinsic occurrences such as loss of services, missile impact, and natural occurrences (e.g., earthquake, tornado). Accident frequencies were determined by formal analysis to be very low. Accident consequences are greatly mitigated by the safety and containment features designed into the fuel modules and shipping cask, the long cooling time since reactor discharge, and the redundant safety features designed into the facilities, equipment, and operating procedures for the PWR Core 2 Project. Possible hazards associated with the handling of these fuels have been considered and adequate safeguards and storage constraints identified. The operations of M-160 cask unloading and module storage will not involve identifiable risks as great or significantly greater than those for comparable licensed nuclear facilities, nor will hazards or risks be significantly different from comparable past 221-T Plant programs. Therefore, it is concluded that the operations required for receipt, handling, and defueling of the M-160 cask and for the storage and surveillance of the PWR Core 2 fuel assemblies at the 221-T Canyon Building can be performed without undue risk to the safety of the involved personnel, the public, the environment or the facility

  14. 14C Behaviour in PWR coolant

    Although 14C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO2), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14C in reactor coolant. A simple chemical kinetic model predicts that CH3OH would be the initial product from radiolytic reactions of 14C following its formation from 17O. CH3OH is predicted to arise as a result of reactions of OH. with CH4 and CH3, and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH3OH can be thermally reduced to CH4 in PWR conditions, although formation of CO2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH4 is the dominant form in PWR and CO2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble material or suspended

  15. Industrywide survey of PWR organics. Final report

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

  16. Minimization of PWR reactor control rods wear

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

  17. Transient study of a PWR pressurizer

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

  18. Determination and declaration of critical nuclide inventories in Belgian NPP radwaste streams

    The nuclear power plants (NPPs) managed by ELECTRABEL are located at the Doel (4 units) and the Tihange (3 units) sites and have a total capacity of 5700 MW(e). All the units are of the PWR type. Taking into account the need for retrievability and reliability of all requested waste data, the operator ELECTRABEL has subcontracted a complete study to the engineering company TRACTEBEL ENERGY ENGINEERING (TEE) in order to elaborate a computer code for the determination of critical nuclides in the different waste streams. This program should guarantee retrievability and reliability of all information related to the waste packages produced at the NPP. Two computer codes, LLWAA and DECL, have therefore been developed by TEE. The first code (LLWAA: low level waste activity assessment code), enables to predict the global inventories and/or the scaling factors of the critical nuclides in the conditioned and in the non-conditioned waste generated by the operation of a PWR. This code is site-specific as it takes into account the plant design characteristics and operating conditions. A version for BWR plants is under development. The second code 'DECL', deals mainly with the complete database management of each waste package produced in order to guarantee full retrievability. LLWAA and DECL are implemented as an integrated software package called 'DECLARE' at the sites of Doel and Tihange. Furthermore, the LLWAA-code has been extended for the determination of the critical nuclides activities in ashes produced by incineration (LLWAA-Ashes) and for the assessment of the critical nuclides activities deposited on equipment of the nuclear auxiliary systems (LLWAA-Decom). (author)

  19. Safety Culture Survey in Krsko NPP

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

  20. Performances of NPP Cernavoda Unit 1 in 1998

    The paper discusses the performances of NPP Cernavoda Unit 1 in 1998, the second year of commercial operation with reference to the electric power production and to other indicators related to nuclear safety, radioprotection, radioactive wastes and nuclear fuel. A comparison with similar indicators reported by other NPP worldwide is presented. The main stages of the operation history as well as some comments on the unplanned events occurred at NPP Cernavoda during 1998 are presented. The NPP operation costs, in-reactor behavior of the nuclear fuel manufactured at 'Nuclearelectrica', Pitesti and the results of the tests for the determination of the specific nuclear fuel consumption are mentioned. (authors)

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

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

  2. Program of upgrading and safety improvement of NPP V-2 units

    This paper dealt with the role of Bohunice NPP in the frame of energy production by Slovak Energy joint company, upgrading and safety improvement of NPP V-2 units, basic documents for defining the goals of NPP V-2 units safety upgrading, assessment of safety of NPP V-2 units, program of upgrading and safety improvement of NPP V-2 units, and with the financial resources necessary for upgrading and safety improvement of NPP V-2 units

  3. Minor actinide transmutation on PWR burnable poison rods

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

  4. CECP, Decommissioning Costs for PWR and BWR

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  5. Navy lifts veil on PWR research

    The author describes the experience of Rolls Royce in developing nuclear reactors for the Navy. Reference is made to the commissioning of HMS Sceptre in February 1978, Britain's 14th nuclear submarine. This event coincided with a decision to lift the veil somewhat on a Research and Development programme that has remained secret for nearly 20 years. Factors that have inhibited progress in this field are mentioned. One of these factors has been the high cost of marine nuclear propulsion systems, tending to limit interest to very large vessels or some special purpose craft. Another factor has been slowness to develop universally acceptable safety criteria, to allow for free and ready access of nuclear vessels to ports. A third factor has been the military origins of much of the development work. A new factor that has arisen recently is the development of the Westinghouse PWR (pressurised water reactor) for marine use in the UK. This has involved collaboration with the US Westinghouse Electric Corporation. Rolls Royce and Associates were chosen to manage this work, which is here described, including the first PWR to be designed and built in Britain and incorporated into a submarine (HMS Vulcan). Much of the design work has been concerned with development of the reactor core and increasing the endurance of the vessel between refuellings. Another aspect was less noise and vibration. Costs of this work are stated, and new test facilities are described. (U.K.)

  6. Workers doses in central European PWR NPPs

    As is stated, the ISOE database which was established in 1992 forms an excellent basis for studies and comparisons of occupational exposure data between nuclear power plants. In the year 2001, 69% of all participating reactors were pressurised water reactors. The ISOE database presents workers' exposure from 213 participating pressurised reactors (PWR) from 27 countries in that year. Among these 32 PWRs belong to six Central European Countries. The analysis of the exposure of workers based on radiation protection performance indicators (collective dose, average dose etc.) in these PWRs could be related to some nuclear safety performance indicators for recent years using ISOE database. The comparison is made to ISOE world - wide data. In the six Central European Countries altogether 32 PWR operated in the year 2001.The international databases of performance indicators related to radiation protection as for example the ISOE or the UNSCEAR database can be use as an efficient tool in the management of radiation protection of workers in a nuclear facilities and regulatory bodies. The databases enable the study of performance trends and the improvement of radiation protection. (authors)

  7. PWR and WWER thorium cycle calculation

    The first step of the investigation of the thorium fuel cycle with HELIOS 1.8 is validation of the results obtained from the code for this particular type of fuel. To complete this first task we performed calculation of the benchmark announced by IAEA in 1995. The benchmark was based on a simplified PWR model of the assembly with reduced fuel composition. This calculation was focused on a comparison of the methods and basic nuclear data. After successful validation of the code we focused our work on calculating the PWR and WWER thorium fuel cycles. The thorium cycle begins after the first use of UO2 fuel in the reactor as separation of plutonium from the burnt fuel. Separated plutonium is mixed with thorium and used as a new nuclear fuel in the reactor. For our calculation we prepared two variants of the assembly - the first variant is a homogeneous distribution and the second one is a non-homogenised distribution of thorium fuel in the assembly. The model of non-homogenised distribution of Pu-Th fuel was designed by replacing selected rods of the classical UO2 assembly by Pu-Th rods. These selected rods are distributed symmetrically in the assembly. Other rods in the assembly remain the same as in the classical UO2 assembly. The calculated and compared values are criticality and fuel composition as a function of burnup (Authors)

  8. Design of NPP of new generation being constructed at the Novovoronezh NPP site

    The design of a new generation NPP is described, underscoring advances in physical attributes and passive safety systems based on experiences with earlier designs at operating NPPs. This paper elaborates on systems for handling and storing radioactive wastes, on refinements in containment measures and on experimental and analytic validation of critical design factors. (author)

  9. The Krsko NPP Analyzer - phase II

    During the assessment of the nuclear safety level in Slovenia the SNSA (Slovenian Nuclear Safety Administration) found a need to gain a software support for analyzing the transients of the Krsko Nuclear Power Plant. Combining the RELAP5 code with graphical interface NPA (Nuclear Plant Analyzer) management staff saw an opportunity to have a powerful instrument for analyses and calculations on a user friendly basis and to get familiar with RELAP5 input deck without extra costs. At the beginning the project started with first phase, where Jozef Stefan Institute - OR4 formed a basic scope of the work and prepared the demo version of the SBLOCA on the basis of RELAP5/Mod3.1code input deck. Tractebel was chosen as a supplier of the project's second phase on the base of public bid. In 1996 the work started with translation of the input model from version Mod2.5 to Mod3.2 with standard routines and small final corrections. After this phase of the project a user of the Krsko NPP Analyzer can run accidents as SBLOCA, Main Steam Line Break, Feedwater Line Break, SGTR, and many other transients activating and combining interactive commands, starting from a full power operation. The third phase is planed. The SNSA has plans to improve the model of the Krsko NPP for a better simulation of core phenomena and to have detailed models of safety and auxiliary systems for increasing the number of possible transients and failures. The Critical Safety Function window will be created as a special mask. The analyzer will be used for education of employees and external experts, who are engaged in case of an emergency, to get familiar with the NPP systems and their operation. During the assessment of the nuclear safety level in Slovenia the SNSA (Slovenian Nuclear Safety Administration) found a need to gain a software support for analyzing the transients of the Krsko Nuclear Power Plant. Combining the RELAP5 code with graphical interface NPA (Nuclear Plant Analyzer) management staff saw an

  10. Corporate portal system at PAKS NPP, Hungary

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

  11. PWR and WWER fuel performance. A comparison of major characteristics

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  12. Factors analysis of water hammer in FLOWMASTER for main feedwater systems of PWR nuclear power plants

    The main feedwater system of a nuclear power plant (NPP) is an important part in ensuring the cooling of a steam generator. It is the main pipe section where water hammers frequently occur. Studying the regulator patterns of water hammers in the main feedwater systems is significant to the stable operation of the system. This article focuses on a parametric study to avoid the consequences of water hammer effect in PWR by employing a general purpose fluid dynamic simulation software-FLOWMASTER. Through FLOWMASTER's transient calculating functions, a mathematical model is established with boundary conditions such as feedwater pumps, control valves, etc., calculations of water hammer pressure when feedwater pumps and control valves shut down, and simulations during instantaneous changes in water hammer pressure. Combining a plethora of engineering practical examples, this research verified the viability of calculating water hammer pressure through FLOWMASTER's transient functions and we found out that, increasing the periods of closure of control valves and feedwater pumps control water hammers effectively. We also found out that changing the intervals of closing signals to feedwater pumps and control valves aid to relieve hydraulic impact. This could be a guideline for practical engineering design and system optimization. (author)

  13. Key issues for the control of refueling outage duration and costs in PWR nuclear power plants

    For several years, EDF, within the framework of the CIDEM project and in collaboration with some German Utilities, has undertaken a detailed review of the operating experience both of its own NPP and of foreign units, in order to improve the performances of future units under design, particularly the French-German European Pressurized Reactor (EPR) project. This review made it possible to identify the key issues allowing to decrease the duration of refueling and maintenance outages. These key issues can be classified in 3 categories: Design; Maintenance and Logistic Support; Outage Management. Most key issues in the design field and some in the logistic support field have been studied and could be integrated into the design of any future PWR unit, as for the EPR project. Some of them could also be adapted to current plants, provided they are feasible and profitable. The organization must be tailored to each country, utility or period: it widely depends on the power production environment, particularly in a deregulation context. (author)

  14. Baltic NPP Project specifics and current status

    Project overview: 2 x 1194 MW Units (AES-2006 series); Location in Kaliningrad region of the; Russian Federation; Operation dates: Unit 1 – Oct 2016; Unit 2 – Apr 2018; Site preparatory works ongoing. This is first NPP project in the Russian Federation providing opportunity for participation of foreign investors. Foreign investors may acquire up to 49% share. Cross-border transmission lines developed under separate project with participation of foreign investors. Conclusion: At the selected set of assumptions, the project is financially feasible in all scenarios

  15. Preparation of NPP Dukovany periodic safety review

    Dukovany NPP in Czech Republic performs a periodic safety review for the second time after approximately 20 years of operation. The history of the Safety Report and its transformation into an internationally accepted form complying with IAEA standards is described. The deterministic and probabilistic assessment of the plant's safety-related design and state is applied to determine whether and to what extend the relevant protective goals are fulfilled by the existing plant design. A description of the step-by-step process is presented together with the creation of methods and criteria for PSR evaluation prepared by Nuclear Research Institute Rez

  16. Mochovce NPP safety improvement and completion

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

  17. Students education and training for Slovak NPP

    Slovak University of Technology is the largest and also the oldest university of technology in Slovakia. Surely more than 50% of high-educated technicians who work nowadays in nuclear industry have graduated from this university. The Department of Nuclear Physics and Technology of the Faculty of Electrical Engineering and Information Technology as a one of seven faculties of this University feels responsibility for proper engineering education and training for Slovak NPP operating staff. The education process is realised via undergraduate (Bc.), graduate (MSc.) and postgraduate (PhD.) study as well as via specialised training courses in a frame of continuous education system. (author)

  18. WWER-type NPP spray ponds screen

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

  19. Nuclear Oversight Function at Krsko NPP

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

  20. Safe 15 Terawatt of Temelin NPP

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

  1. Contact steam condenser for NPP containment shells

    The invention concerns technical and radiation safety of NPP. It enhances the efficiency of the exsisting systems for accident localization or for reduction of the post-accident pressure in the containment shell. The invention consists in stepwise acomplishment of the condenser. The first step, spraying, uses a gravitational vessel of the cooling water inventory with outlet, cooler, water distribution tube with exhaust nozzles. The second step, bubling, is asymmetric and cosists in gate pipe ending with steam-gas mixture distibutor, and gate vessel containing the second part of the cooling water inventory. 3 cls., 1 fig

  2. Optimization of NPP and TPP maintenance programs

    A new structured approach to the maintenance program for NPP and TPP is proposed. The aim is to enhance efficiency by improving nuclear safety, reliability and availability of the plants. This will be achieved by replacing the system of scheduled preventive maintenance with actions of corrective maintenance. The new optimized program includes the following elements: reliability centred maintenance approach; technical diagnostics and forecasting; overall maintenance expenditure reduction; analysis and comparison of data; computerized system for maintenance management with necessary hardware and software. 3 figs., 8 refs

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

    Hirofumi Hashimoto

    2013-02-01

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

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

    2008-01-01

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

  5. Transition cycle fuel management problems of NPP Krsko

    Transition cycle fuel management problems are described and illustrated using results and experience attained during core reload design of NPP Krsko. Improved version of computer code package PSU-LEOPARD/Mcrac is successfully applied to NPP Krsko loading pattern design. (author)

  6. Balancing preventive and corrective maintenance in Cernavoda Unit 1 NPP

    The paper presents a short reminder of Romania's Cernavoda NPP entering commercial operation and a brief description of the CANDU-6 project on which Unit 1 is based. The short term objectives of the maintenance management, the status of the existing maintenance programmes as well as future predictable maintenance programmes are outlined together with the Government plan to complete the balance of NPP. (author)

  7. Unit Commissioning of “Belene” NPP (Bulgaria)

    This presentations gives detailed information about the following topics about commissioning: principles of NPP commissioning; phases of NPP commissioning; organization of commissioning activities; duties and responsibilities of the parties for carrying out unit commissioning activities; responsibility and obligations of the sides during commissioning of power unit; documentation required for power unit commissioning; quality assurance for commissioning activities

  8. Brazil: Angra 2 NPP. Manpower and documentation. Annex 2B

    This annex deals with manpower and documentation. Several initiatives were implemented during the suspension period to preserve manpower and documentation at Angra 2 NPP. They have paid off and construction of the NPP was resumed with relatively little unexpected difficulties. This annex outlines the essentials of these initiatives. (author)

  9. NPP electrical price and tariff in the world

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

  10. Preliminary results of the surveillance extension program of NPP Paks

    The paper describes the preliminary stage of the surveillance extension programme at NPP Paks. To prepare the surveillance extension programme 5 set of specimens including 6 different forging and weldment have been irradiated and tested. The obtained results are used in the final design of the extended surveillance programme of NPP Paks. (author). 3 refs, 6 figs, 1 tab

  11. Sociological investigations on Ignalina NPP and within its surroundings

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

  12. Use of NESTLE computer code for NPP transition process analysis

    A newly created WWER-440 reactor model with use NESTLE code is discussed. Results of 'fast' and 'slow' transition processes based on it are presented. This model was developed for Rovno NPP reactor and it can be used also for WWER-1000 reactor in Zaporozhe NPP

  13. Basic information about development and construction of a PWR

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

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

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

  15. Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation

    Usual way to analyze Main Steam Line Break (MSLB) accident in PWR plants is to calculate core and containment responses in two separate calculations. In first calculation system code is used to address behaviour of nuclear steam supply system and containment is modelled mainly as a boundary condition. In second calculation mass and energy release data are used to perform containment analysis. Coupled code R5G realized by direct explicit coupling of system code RELAP5/MOD3.3 and containment code GOTHIC is able to perform both calculations simultaneously. In this paper R5G is applied to calculation of MSLB accident in large dry containment of NPP Krsko. Standard separate calculation is performed first and then both core and containment responses are compared against corresponding coupled code results. Two versions of GOTHIC code are used, one old ver 3.4e and the last one ver 7.2. As expected, differences between standard procedure and coupled calculations are small. The performed analyses showed that classical uncoupled approach is applicable in case of large dry containment calculation, but that new approach can bring some additional insight in understanding of the transient and that can be used as simple and reliable procedure in performing MSLB calculation without any significant calculation overhead. (author)

  16. Evaluation of tight-pitch PWR cores

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235U/UO2 : Pu/ThO2 : 233U/ThO2 - and the conventional recycle-mode uranium system - 235U/UO2 : Pu/UO2. The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

  17. A pressure drop model for PWR grids

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

    1998-12-31

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

  18. Zebra: An advanced PWR lattice code

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

    2012-07-01

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

  19. Zebra: An advanced PWR lattice code

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

  20. Evaluation model for PWR irradiated fuel

    The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author)

  1. Crevice chemistry control in PWR steam generators

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

  2. The underclad cracking in PWR reactor vessels

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

  3. The material analysis for PWR primary equipment

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

  4. Subcooled decompression analysis in PWR LOCA

    The thermo-hydraulic behavior of the coolant in the primary system of a nuclear reactor is important in the core heat transfer analysis during a hypothetical loss-of-coolant accident (LOCA). The heat transfer correlations are strongly dependent on local thermo-hydraulic conditions of the coolant. The present work allows to calculate such thermo-hydraulic behavior of the coolant during subcooled decompression in PWR LOCA by solving the mass, momentum, and energy conservation equations by the method of characteristics. Detailed studies were made on the transient coolant outflow at the pipe rupture and the effect of frictional loss and heat addition to the coolant on the decompression. Based on the studies, a digital computer code, DEPCO-MULTI, has been prepared and numerical results are compared with the ROSA (JAERI) and the LOFT (NRTS) semiscale test data with various coolant pressures, temperatures, pipe break sizes, and complexity of flow geometry. Good agreement is generally obtained

  5. Recriticality risk in PWR spent fuel pools

    In this paper we investigated the situation in a PWR Spent Fuel Pool (SFP) following a long-term loss of power / loss of cooling accident. In the SFP there is a large amount of water with soluble boron between the fuel assemblies. There may be a problem from the point of view of criticality safety if the water of the SFP starts to boil and evaporate. A thermal-hydraulic analysis was performed using a simplified model of the SFP. The thermal-hydraulic analysis shows that in all cases a chaotic boiling phenomenon develops. This indicates that even if there is an issue of (near-)criticality, it will have a very intermittent nature. The multiplication factor of the SFP was evaluated with a Monte Carlo calculation. The neutronic analysis was performed for several representative cooling situations. In all cases, the system remains (deeply) subcritical. (author)

  6. Exponential experiments on PWR spent fuel assemblies

    An Exponential experiment system which is composed of a neutron detector, a signal analysis system and a neutron source, Cf-252 has been installed in order to experimentally determine the neutron effective multiplication factor for PWR spent fuel assembly. The axial background neutron flux is measured as a preliminary performance test. From the results, the spacer grid position is determined to be consistent with the design specifications within a 2.3% relative error. The induced fission neutron for four of the assemblies is also measured by scanning the neutron source, Cf-252 or the neutron detector. The exponential decay constants have been evaluated by the application of Poisson regression to the net induced fission neutron counts. It was revealed that the average exponential decay constants for the C15, J14, G23 and J44 assemblies were determined to be 0.130, 0.127, 0.125 and 0.121, respectively. (author)

  7. Stochastic optimization of loading pattern for PWR

    The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

  8. Corrective action program at Krsko NPP

    The Krsko NPP develops software that enables electronic reporting of all kind of deviations and suggestions for improvement at the plant. All the employees and permanent subcontractors have the access to the system and can report deviations. NPP has centralized decision process for the distribution of reported deviation. At this point all direct actions are electronically tracked. The immediate benefits of this new tool were: Reporting threshold has been lowered; Number of reporting people has increased; One computerized form for all processes; Decision, which process will solve the deviation, is centralized; All types of deviation are in the same environment; Our experiences of the processes are incorporated in the program; Control of work that has been done; Archiving is electronic only. Software basic data: Application system Corrective action program is a WEB application. Data is stored in Oracle 8.1.7 i database. Users access application through PL/SQL gateway on Oracle 9i Application Server 1.0.2. using Microsoft Internet Explorer browsers(Version 5 or later). Reports are implemented by Oracle Reports 6i. Menus are designed by Apycom Java Menus and Buttons v4.23. Our Presentation will include: Basic idea; Implementation change management; Demonstration of the program.(author)

  9. Central alarm system replacement in NPP Krsko

    Current NPP Krsko central alarm system consists of three main segments. Main Control Board alarm system (BETA 1000), Ventilation Control Board alarm system (BETA 1000) and Electrical Control Board alarm system (BETA 1100). All sections are equipped with specific BetaTone audible alarms and silence, acknowledge as well as test push buttons. The main reason for central alarm system replacement is system obsolescence and problems with maintenance, due to lack of spare parts. Other issue is lack of system redundancy, which could lead to loss of several Alarm Light Boxes in the event of particular power supply failure. Current central alarm system does not provide means of alarm optimization, grouping or prioritization. There are three main options for central alarm system replacement: Conventional annunciator system, hybrid annunciator system and advanced alarm system. Advanced alarm system implementation requires Main Control Board upgrade, integration of process instrumentation and plant process computer as well as long time for replacement. NPP Krsko has decided to implement hybrid alarm system with patchwork approach. The new central alarm system will be stand alone, digital, with advanced filtering and alarm grouping options. Sequence of event recorder will be linked with plant process computer and time synchronized with redundant GPS signal. Advanced functions such as link to plant procedures will be implemented with plant process computer upgrade in outage 2006. Central alarm system replacement is due in outage 2004.(author)

  10. RCM at NPP Dukovany, Czech Republic

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

  11. Forsmark NPP I and C modernization strategy

    By the year 2000, the Forsmark NPP was halfway through the planned plant life. As early as 1995, Forsmark realized that the old analog I and C equipment would need to be replaced before 2005. At the Forsmark NPP they had strength of a vision of an integrated modernization and a strategy to reach the vision. Without vision and strategy, the plant could end up with a fragmented plant I and C-architecture that is not cost-effective or operable. This paper will address several questions that led to the current modernization program in Forsmark, the more important questions are: What would happen if the modernization would be postponed? Which main requirements were to be achieved by means of the modernization strategy? The goal of a completed plant modernization program is a totally integrated system solution and what factors were considered during the modernization? How to gain acceptance from the operational staff in designing Control Room and Soft Control Displays? What are the important roles for the staff and organization to reach the end goal? What has been the experience to date and what are the lessons learned? Thanks to the long term co-operation between Forsmark and Westinghouse the modernization has been very successful for both parties. (orig.)

  12. NPP Krsko Periodic Safety Review action plan

    In the current, internationally accepted, safety philosophy Periodic Safety Reviews (PSRs) are comprehensive reviews aimed at the verification that an operating NPP remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain an acceptable level of safety. These reviews are complementary to the routine and special safety reviews. They are long time-scale reviews intended to deal with the cumulative effects of plant ageing, modifications, operating experience and technical developments, which are not so easily comprehended over the shorter time-scale routine of safety reviews. The review was completed in 2005 and the next period will see the implementation of the action plan including some plant upgrades. The action plan lists issues that should be implemented at NPP Krsko together with associated milestones. The milestones were assumed based on best estimate resource availability and their ends can be potentially floated. In some cases, multiple corrective measures may be postulated to provide resolution for a given safety issue. The Slovenian Nuclear Safety Administration by decree approved the first periodic safety review and the implementation plan of activities arising from it. The entire implementation plan must be carried out by 15 October 2010. Report on the second periodic safety review must be submitted by the NEK not later than 15 December 2013. (author)

  13. NPP unusual events: data, analysis and application

    Subject of the paper are the IAEA cooperative patterns of unusual events data treatment and utilization of the operating safety experience feedback. The Incident Reporting System (IRS) and the Analysis of Safety Significant Event Team (ASSET) are discussed. The IRS methodology in collection, handling, assessment and dissemination of data on NPP unusual events (deviations, incidents and accidents) occurring during operations, surveillance and maintenance is outlined by the reports gathering and issuing practice, the experts assessment procedures and the parameters of the system. After 7 years of existence the IAEA-IRS contains over 1000 reports and receives 1.5-4% of the total information on unusual events. The author considers the reports only as detailed technical 'records' of events requiring assessment. The ASSET approaches implying an in-depth occurrences analysis directed towards level-1 PSA utilization are commented on. The experts evaluated root causes for the reported events and some trends are presented. Generally, internal events due to unexpected paths of water in the nuclear installations, occurrences related to the integrity of the primary heat transport systems, events associated with the engineered safety systems and events involving human factor represent the large groups deserving close attention. Personal recommendations on how to use the events related information use for NPP safety improvement are given. 2 tabs (R.Ts)

  14. Utilization of NPP Krsko plant specific simulator

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

  15. Radiation embrittlement of PWR vessel supports

    Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

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

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

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

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

    2011-03-15

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

  18. The operation experience of SG blowdown demineralizers system in Korea's NPP and the best operating method in PWR

    The steam generator blowdown system(SGBDS)'s are installed to remove the concentrated impurities that are entered with the feedwater or the condensate polishing demineralizer's leakages, to minimize the effluent of radioactive materials in case of the SG's tube rupture and to maintain the lebel of ALARA's guideline in the 10CFR50 App.I. In Korea's NPPs, the SGBDS's design structure and operating procedures are various types although the installing purposes are same. The different designs are the each vessel volume, in series or not, and the resins(cation/anion) mixing ratio, etc. By the analysis results, the best method is the resins mixing ratio is determined by the impurities ratio and the new resin vessel is located after the other vessel to increase the chemical or radioactive material's decontamination factor

  19. Changes in 900 MW PWR alarm processing policy

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs

  20. Characterization of Factors affecting IASCC of PWR Core Internals

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

    2008-09-15

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

  1. Analysis and study on nuclear safety of Mitsubishi PWR

    Theme of safety analysis and study are changing to reflect the needs at the time. This paper introduces the overall aspects of transient and accident analysis performed and presents typical researches related to safety analysis for Mitsubishi PWR. (author)

  2. Hydraulic benchmark data for PWR mixing vane grid

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

  3. Hot Operation of FTL for PWR Fuels Irradiation

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

    2010-05-15

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

  4. Overview of US research related to PWR sump clogging

    In the framework of research of researches related to the PWR sump clogging in Usa, the author presents the history of GSI-191 (assessment of debris accumulation on PWR sump performance), the research to date (technical assessment, regulatory guide and evaluation guidance, model validation), the current and planned tests (chemical effect and calcium silicate tests, latent debris and downstream effect tests, integrated chemical effect tests, EPRI coatings study). (A.L.B.)

  5. Pressure-relieving devices and it's arrangement for PWR

    There are four types of PWR pressure-relieving devices: direct acting safety valve, pilot-operated pressure relief valve, power-operated pressure relief valve and safety valve with auxiliaries. The principle of operation, characteristics, arrangement of the pressure-relieving devices for PWR recently used at home and abroad, confidence of discharge, experience in service and developing trend of the devices are introduced. The first and second type of the devices are emphasised

  6. Post-PKS tailoring steps of a disaccharide-containing polyene NPP in Pseudonocardia autotrophica.

    Hye-Jin Kim

    Full Text Available A novel polyene compound NPP identified in a rare actinomycetes, Pseudonocardia autotrophica KCTC9441, was shown to contain an aglycone identical to nystatin but to harbor a unique di-sugar moiety, mycosaminyl-(α1-4-N-acetyl-glucosamine, which led to higher solubility and reduced hemolytic activity. Although the nppDI was proved to be responsible for the transfer of first polyene sugar, mycosamine in NPP biosynthesis, the gene responsible for the second sugar extending glycosyltransferase (GT as well as NPP post-PKS tailoring mechanism remained unknown. Here, we identified a NPP-specific second sugar extending GT gene named nppY, located at the edge of the NPP biosynthetic gene cluster. Targeted nppY gene deletion and its complementation proved that nppY is indeed responsible for the transfer of second sugar, N-acetyl-glucosamine in NPP biosynthesis. Site-directed mutagenesis on nppY also revealed several amino acid residues critical for NppY GT function. Moreover, a combination of deletions and complementations of two GT genes (nppDI and nppY and one P450 hydroxylase gene (nppL involved in the NPP post-PKS biosynthesis revealed that NPP aglycone is sequentially modified by the two different GTs encoded by nppDI and nppY, respectively, followed by the nppL-driven regio-specific hydroxylation at the NPP C10 position. These results set the stage for the biotechnological application of sugar diversification for the biosynthesis of novel polyene compounds in actinomycetes.

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

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

  8. ATUCHA I NPP - Emergency drill practice

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

  9. Core analysis of the first cycle of Chashma nuclear power plant

    The up coming 300 MWe CHASHMA NPP will provide the opportunity to study the burn-up behavior of the fuel. Our experience is limited to the in-core fuel management studies when fuel burn-up remains within the design limits. The initial core is loaded in three regions with fuel of three different enrichments 2.4 w/o, 2.67 w/o and 3.0 w/o. It is intended to study the enhanced fuel burn-up vis-a-vis the expected cost benefit in due course of time. The core of the Chashma nuclear power plant is that of a typical PWR NPP of 300 MWe capacity. It has 121 fuel assemblies and all of them have identical external dimensions and hydraulic characteristics. The core height is 290 cm and equivalent diameter is 248·6 cm. The core is cooled and moderated by H2O and surrounded by a stainless steel baffle. Each fuel assembly consists of 15x15 rod array and the assembly pitch is 20·03 cm The average discharge burn-up is 30,000 MWd/MTU. Core analysis was carried out for the first cycle at hot full power (HFP). Two dimensional calculations were performed for burn-up analysis including core multiplication, flux distribution, burn-up length, isotopic inventory, peaking factor and critical boron concentration to achieve the economical fuel management within the constraints imposed by safe reactor operation. Calculations indicate that expected burn-up of the first cycle is 13479 MWd/MTU equivalent to 485 EFPD, with 25 ppm of boron is still in the system, which is very near to the design value. Similarly assembly power distribution, pin by pin power distribution and reactivity coefficients, calculated are within the acceptable limits. Efforts are on to improve further these calculations. (author)

  10. IAEA activities on NPP personnel training and qualification

    Activities of IAEA concerning training and qualification of NPP personnel consider the availability of sufficient number of competent personnel which is one of the most critical requirements for safe and reliable NPP operation and maintenance. Competence of personnel is essential for reducing the frequency of events connected to human errors and equipment failures. The IAEA Guidebook on Nuclear Power Plant Personnel Training and its Evaluation incorporates the experience gained worldwide and provides recommendations on the use of SAT being the best practice for attaining and maintaining the qualification and competence of NPP personnel and for quality assurance of training

  11. Industry Operating Experience Process at Krsko NPP

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

  12. Metrological supervision of NPP waste assay system

    A waste assay system of Canberra Industries, Inc. has been recently installed at the Nuclear Power Plant (NPP), Paks, Hungary for the qualification and quantification of low level activity waste up to 100 MBq. In the frame of the metrological supervision of the above system a radioactive certified reference source in drum shape was developed by the OMH for the efficiency calibration of the system. 96 point sources of 152Eu were made and placed equal distances from each other in a similar type of drum used for the waste storage. The activity of one point source was about 0.8 MBq, the total activity of drum was 80 MBq. The average density of material filled into the reference drum was about 0.7 kg/dm3

  13. Environmental qualification program for Krsko NPP

    NEK plant components, including those critical to safe plant operation, deteriorate and wear over service life due to the effects of aging and harsh environmental conditions. Since the plant environment is a source of common-cause failures, an Environmental Qualification (EQ) program is required to ensure and demonstrate the ability of safety-related equipment to perform its design safety function during a design-basis (DBE), even after aging over its service life in the plant. EQ is a requirement for plants licensed by the US NRC, in accordance with 10 CFR 50.49, Regulatory Guide 1.89, NUREG-0588, and IEEE-323. This paper presents the current EQ Program status at Krsko NPP.(author)

  14. Aging assessment of cable for NPP

    Activation energy is measured with UTM (Universal Testing Machine), TGA (Thermo-gravimetric Analyzer) and DMA (Dynamic Mechanical Analyzer) to analyze the aging degree of cables for NPP (Nuclear Power Plant). Insulation power cables containing EPR (Ethylene Propylene Rubber) are arranged for two kinds of specimens which are intact specimens and aged specimens by exposing to LOCA (Loss of Coolant Accident) environmental conditions regulated in IEEE 323. In case of intact specimen, values of activation energy are 1.1 eV for UTM, 1.24 eV with storage modulus and 1.13 eV with loss modulus for DMA, 1.29 eV for TGA, respectively. Damping of specimen under LOCA conditions decreases the activation energy to 0.88 eV for TGA. (author)

  15. NPP Krsko Modelling Using LWRA Code

    The position of 'mixture' and 'two-fluid' thermal-hydraulics models is briefly discussed. Basic characteristics of the LWRA code model and usage are described. LWRA input deck for NPP Krsko, based on standard RELAP5/mod2 nadalization, is developed. Calculation speed during steady state transient was about 10 times greater than real time on Pentium 133 Mhz computer. Steady state results are verified against plant data and against RELAP5/mod2 calculation. Generally speaking good agreement is achieved for this preliminary stage in LWRA nodalization development. The code can be initialised and used in simple and effective way and it will be included in student exercises to demonstrate plant behaviour during operational transients and accidents. (author)

  16. Construction of Belene NPP in Bulgaria

    Presentations concluding remarks: ASE has performed its scope of responsibilities under the Agreement of 29.11.2006 and has achieved great results regarding both the Technical part of the Project and its organization; Though there is a number of unsettled issues under the Project, in particular, the issue related to financing, ASE is willing to continue the Project and works on its development; The Russian Party believes that in case the activities under the Project are continued, Belene NPP will be constructed with high quality and within the time limits prescribed in the Agreement of 29.11.2006: 59 months before Unit 1 take-over into operation and 71 month before Unit 2 take-over into operation, starting from concreting of foundation slab of Unit 1 Reactor building

  17. N-16 monitors: Almaraz NPP experience

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

    1997-02-01

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

  18. Modelling of Krsko NPP using MELCOR

    MELCOR, severe accident analysis tool has been used to form a base for nodalization of Krsko Nuclear Power Plant. There have been number of analysis using other integral computer packages performed in the past including MAAP, and STCP programs as well as mechanistic computer packages such as RELAP5. MELCOR is state of the art program maintained by United States Nuclear Regulatory Commission and is currently used at University of Maribor to assess some of the accident sequences. This contribution presents the database development and its main characteristics, including some of the aspects unused by current analysis and foreseen in the future. Advantages and shortcomings of the database are discussed and future plans including proposed improvements are presented. In addition to the database few test results are presented to show the general behaviour of Krsko NPP simulation. (author)

  19. Shake table investigation at Paks NPP

    In the framework of the earthquake upgrading project at Paks NPP shake table investigations were carried out on cabinets and instrumentation and control instruments. Some devices were tested on newly developed shake tables. In the first series six instrumentation and control cabinets were tested with the main purpose to verify experimentally the results of calculations by COSMOSM code. in the second series the mechanical resistance of 8 instruments was rested. The third series the tested instruments were divided into two groups. Instruments in the first group were tested mechanically, while the functioning of instruments in the second group was tested during shaking. The fourth and fifth series of investigations involve test of new cabinets before installation, their functioning was tested during shaking

  20. Kozloduy NPP - its 20 years operational experience

    The main technical and economic results of 20 year operation of the Kozloduy NPP are summarised. The total electricity produced amounts to more than 200 billions KWh. The energy produced by the plant accounts for more than 40% of the national electricity output since 1993. Details of electricity production in different years by all 6 units are given. The efficiency of units 1-4 (WWER-440) is 29-30% and of units 5-6 (WWER-1000) is 29-33%. The Unit 6 has best characteristics. Since 1989 a safety test is being carried out in cooperation with the IAEA. Units 1 and 2 have been reconstructed in order to enhance safety and reliability. 1 tab

  1. NPP Cernavoda Unit 2 Financing Completion Works

    NPP Cernavoda Unit 2 completion is the highest priority of the Romanian power sector strategy. The nuclear energy represents, through its technological features of adopted solution (a CANDU nuclear power plant) and also through technological and economical performance indicators, the best solution to fulfill the demands concerning the sustainable development and the electricity request. The guidelines of energy strategy regarding the nuclear sector development in Romania are framing in the general policy for energy system development at least costs and they are responding to requests concerning the environment and people protection. The paper presents the financing alternatives for Unit 2 completion works taking into consideration the financing market conditions. The paper presents the impact of the financing conditions on the project efficiency, as well as the facilities offered by the Romanian Government in order to support this project. (author)

  2. Students education and training for Slovak NPP

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

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

    Králik, Juraj

    2013-01-01

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

  4. Nuclear power perspective in China

    China started developing nuclear technology for power generation in the 1970s. A substantial step toward building nuclear power plants was taken as the beginning of 1980 s. The successful constructions and operations of Qinshan - 1 NPP, which was an indigenous PWR design with the capacity of 300 MWe, and Daya Bay NPP, which was an imported twin-unit PWR plant from France with the capacity of 900 MWe each, give impetus to further Chinese nuclear power development. Now there are 8 units with the total capacity of 6100 MWe in operation and 3 units with the total capacity of 2600 MWe under construction. For the sake of meeting the increasing demand for electricity for the sustainable economic development, changing the energy mix and mitigating the environment pollution impact caused by fossil fuel power plant, a near and middle term electrical power development program will be established soon. It is preliminarily predicted that the total power installation capacity will be 750-800GWe by the year 2020. The nuclear share will account for at least 4.0-4.5 percent of the total. This situation leaves the Chinese nuclear power industry with a good opportunity but also a great challenge. A practical nuclear power program and a consistent policy and strategy for future nuclear power development will be carefully prepared and implemented so as to maintain the nuclear power industry to be healthfully developed. (author)

  5. Training change control process at Cernavoda NPP

    Full text: The paper presents the process of 'Training Change Control' at Cernavoda NPP. This process is a systematic approach that allows determination of the most effective training and/or non-training solutions for challenges that may influence the content and conditions for a training program or course. Changes may be the result of: - response to station systems or equipment modifications; - new or revised procedures; - regulatory requirements; - external organizations requirements; - internal evaluations meaning feedback from trainees, trainers, management or post-training evaluations; - self-assessments; - station condition reports; - operating experience (OPEX); - modifications of job scope; - management input. The Training Change Control Process at Cernavoda NPP includes the following aspects. The first step is the identification of all the initiating factors for a potential training change. Then, retain only those, which could have an impact on training and classify them in two categories: as deficiencies or as enhancement suggestions. The process is different for the two categories. The deficiency category supposes the application of the Training Needs Analysis (TNA) process. This is a performance-oriented process, resulting in more competent employees, solving existing and potential performance problems. By using needs analysis to systematically determine what people or courses and programs are expected to do and gathering data to reveal what they are really doing, we can receive a clear picture of the problem and then we can establish corrective action plans to fix it. The process is supported by plant subjects matter and by training specialists. On the other hand, enhancements suggestions are assessed by designated experienced persons and then are implemented in the training process. Regarding these two types of initiating factors for the training change control process, the final result consists of a training improvement, raising the effectiveness

  6. Detection of primary coolant leaks in NPP

    The thermo-hydraulic analyses of the SG box behaviour of Kozloduy NPP units 3 and 4 in case of small primary circuit leaks and during normal operation of the existing ventilation systems in order to determine the detectable leakages from the primary circuit by analysing different parameters used for the purposes of 'Leak before break' concept, performed by ENPRO Consult Ltd. are presented. The following methods for leak detection: measurement of relative air humidity in SG box (can be used for detection of leaks with flow rate 3.78 l/min within one hour at ambient parameters - temperature 400 - 600C and relative humidity form 30% to 60%); measurement of water level in SG box sumps (can not be used for reliable detection of small primary circuit leakages with flow rate about 3.78 l/min); measurement of gaseous radioactivity in SG box( can be used as a general global indication for detection of small leakages from the primary circuit); measurement of condensate flow after the air coolers of P-1 venting system (can be used for primary circuit leak detection) are considered. For determination of the confinement behaviour, a model used with computer code MELCOR has been developed by ENPRO Consult Ltd. A brief summary based on the capabilities of the different methods of leak detection, from the point of view of the applicability of a particular method is given. For both Units 3 and 4 of Kozloduy NPP a qualified complex system for small leak detection is planned to be constructed. Such a system has to unite the following systems: acoustic system for leak detection 'ALUS'; system for control of the tightness of the main primary circuit pipelines by monitoring the local humidity; system for primary circuit leakage detection by measuring condensate run-off in collecting tank after ventilation system P-1 air coolers

  7. Digital Components in Swedish NPP Power Systems

    Swedish nuclear power plants have over the last 20 years of operation modernised or exchanged several systems and components of the electrical power system. Within these works, new components based on digital technology have been employed in order to realize functionality that was previously achieved by using electro-mechanical or analogue technology. Components and systems such as relay protection, rectifiers, inverters, variable speed drives and diesel-generator sets are today equipped with digital components. Several of the systems and components fulfil functions with a safety-role in the NPP. Recently, however, a number of incidents have occurred which highlight deficiencies in the design or HMI of the equipment, which warrants questions whether there are generic problems with some applications of digital components that needs to be addressed. The use of digital components has presented cost effective solutions, or even the only available solution on the market enabling a modernisation. The vast majority of systems using digital components have been operating without problems and often contribute to improved safety but the challenge of non-detectable, or non-identifiable, failure modes remain. In this paper, the extent to which digital components are used in Swedish NPP power systems will be presented including a description of typical applications. Based on data from maintenance records and fault reports, as well as interviews with designers and maintenance personnel, the main areas where problems have been encountered and where possible risks have been identified will be described. The paper intends to investigate any 'tell-tales' that could give signals of unwanted behaviour. Furthermore, particular benefits experienced by using digital components will be highlighted. The paper will also discuss the safety relevance of these findings and suggest measures to improve safety in the application of digital components in power systems. (authors)

  8. Indicators to monitor NPP operational safety performance

    Since December 1995 the IAEA activities on safety performance indicators focused on the elaboration of a framework for the establishment of an operational safety performance indicator programme. The development of this framework began with the consideration of the concept of NPP operational safety performance and the identification of operational safety attributes. For each operational safety attribute, overall indicators, envisioned as providing an overall evaluation of relevant aspects of safety performance, were established. Associated with each overall indicator is a level of strategic indicators intended to provide a bridge from overall to specific indicators. Finally each strategic indicator was supported by a set of specific indicators, which represent quantifiable measures of performance. The programme development was enhanced by pilot plant studies, conducted over a 15 month period from January 1998 to March 1999. The result of all this work is compiled in the IAEA-TECDOC-1141, to be published shortly. This paper presents a summary of this IAEA TECDOC. It describes the operational safety performance indicator framework proposed and discusses the results of and lessons learned from the pilot studies. Despite the efforts described, it is clear that additional research is still necessary in areas such as plant-specific adaptation of proposed frameworks in order to suit individual data collection systems and plant characteristics, indicator selection, indicator definition, goal setting, action thresholds, analysis of trends, indicator display systems, analysis of overall safety performance (i.e., aggregation or combination of indicators), safety culture indicators, qualitative indicators, and use of additional indicators to address issues such as industrial safety attitude and performance, staff welfare, and environmental compliance. This is the rationale for a new IAEA Coordinated Research Project on 'Development and application of indicators to monitor NPP

  9. Plant performance monitoring program at Krsko NPP

    A high level of nuclear safety and plant reliability results from the complex interaction of a good design, operational safety and human performance. This is the reason for establishing a set of operational plant safety performance indicators, to enable monitoring of both plant performance and progress. Performance indicators are also used for setting challenging targets and goals for improvement, to gain additional perspective on performance relative to other plants and to provide an indication of a potential need to adjust priorities and resources to achieve improved overall plant performance. A specific indicator trend over a certain period can provide an early warning to plant management to evaluate the causes behind the observed changes. In addition to monitoring the changes and trends, it is also necessary to compare the indicators with identified targets and goals to evaluate performance strengths and weaknesses. Plant Performance Monitoring Program at Krsko NPP defines and ensures consistent collection, processing, analysis and use of predefined relevant plant operational data, providing a quantitative indication of nuclear power plant performance. When the program was developed, the conceptual framework described in IAEA TECDOC-1141 Operational Safety Performance Indicators for Nuclear Power Plants was used as its basis in order to secure that a reasonable set of quantitative indications of operational safety performance would be established. Safe, conservative, cautious and reliable operation of the Krsko NPP is a common goal for all plant personnel. It is provided by continuous assurance of both health and safety of the public and employees according to the plant policy stated in program MD-1 Notranje usmeritve in cilji NEK, which is the top plant program. Establishing a program of monitoring and assessing operational plant safety performance indicators represents effective safety culture of plant personnel.(author)

  10. What is Tacit Knowledge in NPP Maintenance?

    Nuclear power plants have recognized the importance of preserving critical knowledge due to many nuclear experts retiring simultaneously in the near future. The different characteristics of explicit and tacit knowledge have implications on how these types of knowledge can and should be preserved and transferred from the retiring experts to the novices in the NPPs. This paper provides a literature review of the characteristics of tacit knowledge. In addition the DIAMOND model of tacit knowledge in NPP maintenance is presented. In conclusion: In addition to knowing how to perform maintenance tasks (having procedural tacit knowledge), the maintenance workers have to understand the 'what' and 'why' perspectives of the work, e.g. have appropriate cognitive tacit knowledge on the NPP systems, processes and devices as well as knowledge on the reasons and backgrounds for different behaviors. We need to take into account the various aspects of the tacit knowledge when we are determining the methods and aims for sharing tacit knowledge between the employees and especially between experts and novices. Sharing tacit knowledge is not a new invention but observation and imitation (i.e. apprenticeship) is probably the oldest method for teaching and learning. The current emphasis on workplace learning, however, further emphasizes this and the idea that people should become more aware and reflective, and hence, potentially more in control of what and how they learn (Alred and Garvey, 2000). Therefore, as learners and experts are not passive but actively shape their cognitive and procedural tacit knowledge (mainly unconsciously), we need to actively involve them into designing employee training programs and competence development efforts. (authors)

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

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

    1995-08-01

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

  12. Seawater desalination using reusable type small PWR

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

    2003-07-01

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

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

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

    2014-01-01

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

  14. Computer security of NPP instrumentation and control systems: regulatory framework

    The paper examines the regulatory framework on computer security of NPP instrumentation and control systems (I and C) and presents the short overview of IAEA Nuclear Security Series. It considers the key reference manual from these series and draft new guide on NPP I and C computer security. The paper presents requirements for information and computer security of NPP I and C from the standards of the International Electrotechnical Commission (IEC) and, in particular, the standard regulating requirements for NPP I and C computer security program. Regulatory guide of the US Nuclear Regulatory Commission with requirements for computer security program of nuclear facilities has been analyzed. The research considers challenges of regulatory control in this area and defines tasks to improve regulatory framework on computer security at Ukrainian nuclear facilities.

  15. Role of quality management system in ensuring of NPP safety

    The article describes some system factors of the accident on Chornobyl NPP in 1986. The analysis shows that we can improve the safety of nuclear power plants through the implementation of integrated management systems.

  16. Appendix XVII: Knowledge dissemination for NPP senior managers

    Proneos GmbH is an Innovation Management Consultancy providing advisory services to energy and utility corporations, with a specialization on Nuclear Knowledge Management (NKM). The material presented in this appendix is related to work with a leading European Utility Corporation operating several NPPs. A pivotal element of KM in NPP, and a crucial prerequisite for its success, is the proactive support of knowledge dissemination by senior management in technical, commercial and legal domains. The practice described in this appendix is targeting this proactive support through demonstration of KM benefits in a workshop with NPP senior managers. This practice is applicable to organizations operating several NPP and experiencing technical, process-related, organizational and/or cultural challenges of inter-NPP knowledge exchange

  17. BATAN Contribution for NPP-Radioactive Waste Management Scenario

    BATAN contribution for NPP-radioactive waste management scenario has been studied. The study analysis base on the knowledge background, about relation between NPP and NPP-radioactive waste management installation obtained from literatures. The study concluded that BATAN contribution for NPP-radioactive waste management have 5 possibility scenario, which are: only prepare and operate ultimate waste disposal; prepare and operate ultimate waste disposal and interim storage; prepare and operate ultimate waste disposal, interim storage and treat solid radioactive waste by incineration process, prepare and operate ultimate waste disposal and interim storage and treat all of solid radioactive waste; prepare and operate ultimate waste disposal and interim storage and all of radioactive waste. (author)

  18. Introduction of the SAT based training programs at Paks NPP

    An introduction of the SAT based training programs at Paks nuclear power plant is described in detail, including framework of project operation; project implementation; process of SAT applied at Paks NPP and the needs of its introduction

  19. Second loop performance analysis of Cernavoda NPP Unit 1

    This study aims at simulating the thermal process in the second loop of Cernavoda NPP Unit 1 and analyzing the improvement possibilities of the second loop and NPP's performances by using the CYCLE TEMPO software. At the same time, the software product for the simulation of the second loop of Cernavoda NPP can be a useful tool for operation personnel training. In this respect, this paper can be considered as a doorway into a study on creating a software able to simulate the specific steady state regimes of NPP and specific emerging thermodynamic mis-function of certain components in the secondary loop. A conclusion of this paper is that in the frame of the original design, by keeping the same equipment, there are only minor possibilities for efficiency improvement. (authors)

  20. Codes for NPP severe accident simulation: development, validation and applications

    The software tools that describe various safety aspects of NPP with VVER reactor have been developed at the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN). Functionally, the codes can be divided into two groups: the calculation codes that describe separate elements of NPP equipment and/or a group of processes and integrated software systems that allow solving the tasks of the NPP safety assessment in coupled formulation. In particular, IBRAE RAN in cooperation with the nuclear industry organizations has developed the integrated software package SOCRAT designed to analyze the behavior of NPP with VVER at various stages of beyond-design-basis accidents, including the stages of reactor core degradation and long-term melt retention in a core catcher. The general information about development, validation and applications of SOCRAT code is presented and discussed in the paper. (author)

  1. Human reliability analysis for accident sequences in NPP

    The purpose of this paper is to perform a human performance analysis in accident conditions for the operating NPP. This analysis is realized using Human Reliability Analysis (HRA) methods. HRA methods have necessary tools to analyze the human actions, to estimate the human error probabilities and to identify the major factors which could have a negative influence on the mitigating of the consequences of the abnormal events in NPP. The analyzed events are from CANDU 600 NPP. In order to achieve the analysis of these events the THEP and SPAR-H methods were used. After analyzing the results the actuated equipment, the negative influence factors on the human performance and the dependence levels between the human actions and between the human actions and diagnosis were established. In addition, some recommendations were formulated which could influence positive the human performance on the mitigating of the consequences of the accident sequences in NPP. (authors)

  2. Westinghouse advanced passive 600 MWe PWR design

    Although there has been a sharp downturn in the ordering of commercial nuclear power plants throughout the world, it is nonetheless anticipated that this form of energy will remain vital to the economy of many nations in a long term. One of the important new development activities is that of small plants incorporating passive safety features. The small plants have the merits in terms of low total capital requirement and potentially short lead time. The Electric Power Research Institute sponsored the development of an advanced LWR plant in a nine month Westinghouse program, which terminated in March, 1986. Further development at Westinghouse is now in progress on this design called AP 600 under the sponsorship of the U.S. Department of Energy. On the basis of the proven 600 MWe PWR plant design, the specific design improvement for increased safety and operational margin, reduced plant capital and operating cost, simplified plant systems and components, and increased certainty of meeting construction schedule and cost is pursued. The Westinghouse two-loop plants are very competitive, and the operating performance is outstanding by the comparison of plant capacity factor. The operation and maintenance costs are low. The specific design and the features of modification and improvement are discussed. (Kako, I)

  3. Conceptual study of advanced PWR core design

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

  4. Computer aided information system for a PWR

    The computer aided information system (CAIS) is designed with a view to improve the performance of the operator. CAIS assists the plant operator in an advisory and support role, thereby reducing the workload level and potential human errors. The CAIS as explained here has been designed for a PWR type KLT- 40 used in Floating Nuclear Power Stations (FNPS). However the underlying philosophy evolved in designing the CAIS can be suitably adopted for other type of nuclear power plants too (BWR, PHWR). Operator information is divided into three broad categories: a) continuously available information b) automatically available information and c) on demand information. Two in number touch screens are provided on the main control panel. One is earmarked for continuously available information and the other is dedicated for automatically available information. Both the screens can be used at the operator's discretion for on-demand information. Automatically available information screen overrides the on-demand information screens. In addition to the above, CAIS has the features of event sequence recording, disturbance recording and information documentation. CAIS design ensures that the operator is not overburdened with excess and unnecessary information, but at the same time adequate and well formatted information is available. (author). 5 refs., 4 figs

  5. Aging effects in PWR power plants components

    Borges, Diogo da S.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

  6. Enriched Gadolinium as burnable absorber for PWR

    This paper is a summary of a master of thesis work in reactor physics made by Ola Seveborn. The work was done at Vattenfall Braensle AB and Ola was guided through the work by the corresponding author of this paper. The results presented are calculations for Ringhals 3, which is a Westinghouse 3-loop PWR within the Vattenfall Group. The fuel is characterized by 17x17 assemblies of AFA type containing 3.80-3.95 w/o 235U and 8 rods containing 2 w/o Gadolinium with an enrichment of 70 w/o 157Gd. The calculations were performed with the Studsvik-Scandpower code package based on the CASMO-4 lattice code and the SIMULATE-3 nodal code. The results are compared to the corresponding calculations for fuel with 5 w/o gadolinium with natural isotopic constitution. The depletion of the cores was done separately for the reference and enriched case. The results show that the gains in average for the five cycles studied are about 70 EFPH per cycle. This is an effect of the lower gadolinium content needed. Also less parasitic absorption of enriched gadolinium in the end of the fuel life contributes to the increased cycle lengths. The abruptly increased reactivity and internal power peaking factor around 10 MWd/kgU do not affect the core design negatively. (authors)

  7. Maintenance technologies for SCC of PWR

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

  8. Conceptual study on advanced PWR system

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

  9. PWR core monitoring system and benchmarking

    The PWR Power Shape Monitoring System (PSMS) provides site engineers with new capabilities for monitoring and predicting core power distributions. These capabilities can lead to increased plant output as a result of greater operating margins, better load maneuvering, earlier detection of anomalies, and improved fuel reliability. The heart of the PSMS consists of nodal code (NODEP-2/THERM-P) that computes the 3-D core power distribution. This code is coupled to a simplified nodal version of the COBRA-IIIC/MIT-2 thermal-hydraulic model to determine the DNBR. These calculations can be completed in about 30 seconds on a PRIME-750 mini computer. Activation of the calculations and review of the results is through user-friendly interactive software that can be tailored to the requirements and capabilities of the different categories of users through table-driven menus. The PSMS provides unique advances over core power monitoring systems based purely on measurements. The PSMS approach permits the three-dimensional core simulation model to be routinely corrected with in-core/ex-core measurements while simultaneously identifying consistent instrument errors

  10. Tritium management in PWR fuel reprocessing plants

    Activity, quantity and nature of tritium compounds obtained during head end process (cutting and dissolution) are determined to estimate environmental release hazards in fuel reprocessing plants. Measurements on representative PWR reactor fuels (burnup 33,000 MWdt-1, specific power 30 MW dt-1) show that about 60% of the tritium produced in the reactor diffuses in the cladding where it is fixed. Remaining tritium stays in the irradiated oxide and is found as tritiated water in the solution obtained during fuel dissolution. In the UP3 plant at La Hague (France) tritiated water is disposed into the sea without environmental problems. In the case of a reprocessing plant far from the sea, the PUREX process is slightly modified for concentration of tritium in a limited amount of water (TRILEX process). It is verified experimentally in αβγ lab on actual fuel and by simulation at the pilot seale that the supplementary step ''tritium washing'' of the solvent can be obtained in pulsed columns. 4 tables, 7 figs

  11. Modeling of PWR fuel at extended burnup

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  12. Scaling Analysis for PWR Steam Generator

    Li, Yuquan [State Nuclear Power Technology R and D Center, Beijing (China); Ye, Zishen [Tsinghua University, Beijing (China)

    2011-08-15

    To test the nuclear power plant safety system performance and verify the relative safety analysis code, a widely used approach is to design and construct a scaled model based on a scaling methodology. For a pressurized water reactor (PWR), the SG scaling analysis is important before designing a scale model which is expected to well simulate the system response to the accident. In this work, a review of the transient process in SG during a loss of coolant accident (LOCA) is first presented, and then a brief natural circulation scaling analysis is performed to get the basic SG scaling design rules. The U-tube scaling design shows the scaling will enlarge the thermal center height ratio while keeping the length ratio when the scale model uses a different diameter ratio and the height ratio, which causes distortion in natural circulation simulation. And then, by the heat transfer scaling analysis, a relation between the U-tube diameter ratio and model height ratio is obtained, and it shows the diameter ratio decreases with the decreasing model height ratio. In the end, the SG transition from the heat sink to the heat source is analyzed, and the results show the SG secondary inventory and the total material heat capacity need to be properly scaled to represent the transition correctly.

  13. Scaling Analysis for PWR Steam Generator

    To test the nuclear power plant safety system performance and verify the relative safety analysis code, a widely used approach is to design and construct a scaled model based on a scaling methodology. For a pressurized water reactor (PWR), the SG scaling analysis is important before designing a scale model which is expected to well simulate the system response to the accident. In this work, a review of the transient process in SG during a loss of coolant accident (LOCA) is first presented, and then a brief natural circulation scaling analysis is performed to get the basic SG scaling design rules. The U-tube scaling design shows the scaling will enlarge the thermal center height ratio while keeping the length ratio when the scale model uses a different diameter ratio and the height ratio, which causes distortion in natural circulation simulation. And then, by the heat transfer scaling analysis, a relation between the U-tube diameter ratio and model height ratio is obtained, and it shows the diameter ratio decreases with the decreasing model height ratio. In the end, the SG transition from the heat sink to the heat source is analyzed, and the results show the SG secondary inventory and the total material heat capacity need to be properly scaled to represent the transition correctly

  14. Analysis of reactivity accidents in PWR'S

    This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

  15. Conceptual study of advanced PWR core design

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

    1997-09-01

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

  16. Aging effects in PWR power plants components

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

  17. The Regulatory System for the First NPP in Indonesia

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

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

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

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

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

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

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

  1. Quantitative analysis of psychological personality for NPP operators

    The author introduces the relevant personality quantitative psychological research work carried out by 'Prognoz' Laboratory and Taiwan, and presents the primary results of the research for Chinese Nuclear Power Plant (NPP) operator's psychological personality assessment, which based on the survey of MMPI, and presents the main contents for the personality quantitative psychological research in NPP of China. And emphasizes the need to carry out psychological selection and training in nuclear industry

  2. Seismic and geological conditions at the Bohunice NPP site

    The paper brings basic information on geological and seismic characteristics of the site of NPP Jaslovske Bohunice, Slovakia. Western Carpathians and Trnava, bay geological properties are briefly introduced. The most important macroseismic data and data obtained from field measurements are analysed. Main features of the expected strong seismic motion are discussed. The attention is devoted to local soil characteristics just under the site of NPP. (author)

  3. Mochovce NPP unit 1 and 2 completion project

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

  4. Research for techniques in NPP safety software verification and validation

    Digital instrument and control system is main phase of development in nuclear power station. Comparing to analog system, digital I and C system has various advantages and can fulfill function by software which is limited to analog I and C system. This paper introduces profile of NPP I and C system, analyzing safety system and safety software, at last, attention must be paid to safety software verification and validation techniques so as to comply with NPP requirement. (author)

  5. Computerized systems of NPP operators support. (Psychological problems)

    Operator psychological problems arising in the work with NPP operators support computerized systems (OSCS) are considered. The conclusion is made that the OSCS intellectual application will bring the operator into dangerous dependence on his computerized assistant. To avoid this danger it is necessary by creation of the OSCS to divide specially the tasks areas of the operator and OSCS in order to assure the active role of the operator in the NPP control

  6. Safety Improvement at the Ignalina NPP (SIP-2)

    Implementation of SIP-2 began in 1997. The programme envisages measures through introduction of which conclusions concerning safety improvement given in SAR-1 are implemented. Implementation of SIP-2 measures is a precondition for the validity of the operating licence for Unit 1 at the Ignalina NPP, and for issuing the operating licence for Unit 2, to whose licensing the Ignalina NPP and VATESI are getting ready. Results of the Programme are described

  7. Condition of safety protection at the Zaporozhe NPP

    In this paper the radiation conditions at the Zaporozhe NPP are presented. Data for individual and collective personnel radiation doses are given. The dry spent fuel storage situated at the plant site and the activities related to its commissioning are highlighted. Some results from the environmental radiation monitoring of the Zaporozhe NPP area are presented. An assessment of the annual dose from the radioactive releases is made

  8. External costs of nuclear cycles: the case of Ignalina NPP

    The paper discuss external costs of nuclear fuel cycles and applies simplified compact pathway approach developed by IAEA for the evaluation of external costs associated with Ignalina NPP. External energy costs and their evaluation techniques for nuclear and other fuel cycles are analyzed and compared. evaluation of external costs for different fuel cycles was performed for Lithuanian energy sector seeking to evaluate environmental impact due to increased atmospheric pollution because of Ignalina NPP closure

  9. Morbidity with temporary disability in Kozloduy NPP workers

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

  10. Human Resources Training Requirement on NPP Operation and Maintenance

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

  11. Compatibility of Muria NPP Development Planning with Conservation Area

    Compatibility of NPP development planning at Ujung Lemahabang (ULA) site located at Balong Village, Kembang District, Jepara Regency have been assessed of its compatibility with conservation area that are written down at National Spatial Development Planning (RT/RW National), Province RT/RW and Regency RT/RW. In the context of NPP development planning, conservation area should be considered as one of significant aspect in accordance with the regional spatial planning. The objective of the assessment is to analyze the compatibility of NPP site with area conservation located at surrounding the site, and also to provide safety guarantee to conservation area from radiation hazard both in the construction and operation stage including postulated accident. Based on the concept, NPP development planning should be refers to spatial development planning that will assure of the conservation existence therefore it will be support countermeasure if there is any accident. The method of the research used were identification of all conservation area, zoning of each conservation area surrounding the site, furthermore analyzing compatibility of the site to the conservation area. The result of the research showed that Muria NPP site was suitable with area conservation surrounding the NPP Site. (author)

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

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

  13. Seismic qualification of PWR plant auxiliary feedwater systems

    Lu, S.C.; Tsai, N.C.

    1983-08-01

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

  14. Seismic qualification of PWR plant auxiliary feedwater systems

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

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

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

  16. The development of flow test technology for PWR fuel assemblies

    The objective of this project is to design and construct a high temperature and pressure flow test facility and to develop flow test technology for the evaluation of PWR fuel performance. For the nuclear fuel safety aspect it is of importance to evaluate the thermalhydraulic compatibility and mechanical integrity of a newly designed fuel through the design verification test. The PWR-Hot Test Loop facility is under construction to be used to perform a pressure drop test, a lift force test and a fretting corrosion test of a fullsize PWR fuel assembly at reactor operating conditions. This facility was designed to be used to produce the hydraulic parameters of the existing PWR fuel assemblies(14x14FA, 16x16FA, 17x17FA) and to verify a design of advanced fuel assemblies (KAFA-I and KAFA-II) developed by KAERI. The PWR-Cold Test Loop facility with the 5x5 Rod Bundles in the test section was designed and installed to carry out the flow distribution study by means of Laser Doppler Velocimeter. The LDV techniques have been developed and used to measure the flow velocity and turbulent intensity for evaluating mixing effects of a newly designed spacer grid with and without mixing vanes, cross flow between the fuel assemblies and a turbulent model. (Author)

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

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

    2013-10-15

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

  18. Students education and training for Slovak NPP

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

  19. Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5

    Highlights: • A PWR spent fuel pool was modeled by using MAAP5. • Loss of pool cooling severe accident scenarios were studied. • Loss of pool cooling accidents with two mitigation measures were analyzed. - Abstract: The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents the analysis of loss of pool cooling accident scenarios and the discussion of mitigation measures for the SFP at a pressurized water reactor (PWR) NPP with the MAAP5 code. Analysis of uncompensated loss of water due to the loss of pool cooling with different initial pool water levels of 12.2 m (designated as a reference case) and 10.7 m have been performed based on a MAAP5 input model. Scenarios of the accident such as overheating of uncovered fuel assemblies, oxidation of claddings and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission products were predicted with the assumption that mitigation measures were unavailable. The results covered a broad spectrum of severe accident evaluations in the SFP. Furthermore, as important mitigation measures, the effects of recovering the SFP cooling system and makeup water in SFP on the accident progressions have also been investigated respectively based on the events of pool water boiling and spent fuels uncovery. Based upon the reference case, three cases with the recovery of SFP cooling system and three other cases with makeup water in SFP have been studied. The results showed that, severe accident might happen if SFP cooling system was not restored timely before the spent fuels started to become uncovered; spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate larger than the average evaporation rate defined as the division of pool water mass above the

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

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

  1. Stade NPP. Dismantling of the reactor pool

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

    2012-11-01

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

  2. EPC projects for EPR Flamanville 3 NPP

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

  3. EPC projects for EPR Flamanville 3 NPP

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

    2010-07-01

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

  4. Barsebaeck NPP in Sweden - Decommissioning Project

    Barsebaeck 1 and 2, type BWR (Boiling Water Reactor) with a capacity of 615 MWe was closed down permanently on 30 November 1999 respective 31 May 2005 due to political decision. Both units together have been in Service operation (Care and maintenance) since 1 December 2006. Barsebaeck NPP will stay in Service operation until beginning of 2018 when Dismantling operation begins with the aim of a free-realized site in the beginning of 2025. That means that the remaining buildings, including equipment should be declared free-released or dismantled. It would then be up to the owner, E.ON Kaernkraft Sverige AB (EKS) to decide what is to be done with the site in the future. There was a re-organisation at Barsebaeck Kraft AB (BKAB) in 1 January 2007 and the company is organised in the following areas of function: site service operation, decommissioning planning, new business and BO replacement. The Organisation at BKAB has gone down from 450 during operation of Barsebaeck 1 and 2 to 50 employees (2009-01-01) involved in Service operation of both units. But still there are in total 250 persons placed at Barsebaeck NPP with different kinds of job assignments. A lot of activities have been carried out since 2000 and up to now for example: - All nuclear fuel has been transported away to interim storage at CLAB in Oskarshamn. - BKAB have built up contact nets and competence by taking part in different kinds of national and international organisations (SKB, IAEA, OECD/NEA TAG, WNA, ENISS, WANO, EPRI etc) commissions. - The Electrical and operational systems have been rebuilt for the actual demands and requirements for the Service operation. - The central control room is unattended since 17 December 2007 and the supervision of the Service operation is handled by a system of VDI (duty engineers) and LOP (alarm operators). - Full system decontamination on unit 1 and 2. Barsebaeck's approach today and for the future dismantling are: - Safer; - Faster; - Cost effective. BKAB

  5. Stade NPP. Dismantling of the reactor pool

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

  6. Experience of Bohunice V-1 NPP

    Slovakia remains significantly dependent on imports of primary energy sources, which represent as much as 80% of the demand. Of the total consumption of electricity in Slovakia, 40% was generated in nuclear power plant units in 1998. Slovakia operates 6 units with WWER 440 nuclear reactors. Slovakia is the signatory of all important international agreements and conventions in the field of nuclear energy, and its legislation is in an advanced stage of approximation to European Union law. This is a very important aspect, showing Slovakia's approach to nuclear safety. In 1993 Slovakia accepted the commitments of the UN Convention on Climate Changes, including a reduction of greenhouse gases to 1990 levels by the year 2000. Moreover, as an internal target Slovakia has set the reaching of the 'Toronto Objective', i.e. 20% reduction in COx, emissions through the year 2005 as compared to 1988. In our opinion, this is not possible without nuclear energy. Time has shown, that the political aspects are more powerful, especially if you underestimate their importance over the than the technical ones. In the case of Bohunice V-1 NPP the political aspects were on the following levels: 1. Slovak republic (Czechoslovakia), political changes, decisions of the government; 2. European Union - Agenda 2000, Accession criteria, nuclear safety criteria, EBRD; 3. Austria as a neighbouring country. Starting with year 1990, 23 expert missions took place at Bohunice V-1 NPP by now. The only criteria for further operation should have been Nuclear safety, which is supervised by NRA SR. It was fully in compliance with EU policy, each country is solely responsible for its energy sector and for nuclear energy use. Our satisfaction lasted not too long. Following negotiation with EU on the highest political level, driven by willingness to be invited for negotiation of accession on the Helsinki Summit, the Slovak government decided on September 14th, on Bohunice V-1 Units shutdown in 2006 and 2008

  7. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

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

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

  9. Signal processing methods for PWR reactor noise diagnostic system

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

  10. PWR fuel performance and future trend in Japan

    Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of RandD on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important RandD targets are the burnup extension, Gd contained fuel, utilization and the load follow capability

  11. Fabrication of PWR fuel assembly and CANDU fuel bundle

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  12. Timing analysis of PWR fuel pin failures

    This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PF1/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design burnup. Using peaking factors commensurate with actual burnups would result in longer intervals for both reactor designs. This document provides appendices K and L of this report which provide plots for the timing analysis of PWR fuel pin failures for Oconee and Seabrook respectively

  13. Load-following operation of PWR plants

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

  14. Load-following operation of PWR plants

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

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

  15. Borssele PWR noise: measurements, analysis and interpretation

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

  16. The safety of Ignalina NPP and ecological danger in public opinion of inhabitants of Daugavpils region

    Inquest of Daugavpils' region pointed to a big anxiety for ecological danger of Ignalina NPP by inhabitants and experts. Absolute majority of respondents (73-78% inhabitants and 68-82% experts) apprehend the NPP as very dangerous and dangerous. More than half of respondents apprehend the dangerous increased during last two-three years. It is because no one has a good reference about situation, because tragic al Chernobyl NPP burst was on. The anxiety increases if the respondent lives nearer of NPP. Inhabitants of Daugavpils and it's region wants the better reference about situation, about future of Ignalina NPP after 2010 year, about securities means in case of NPP burst. (author)

  17. Ukrainian Nuclear Society International Conference 'Modernization of the NPP with VVER reactor' (abstracts)

    Abstracts of the papers presented at International conference of the Ukrainian Nuclear Society 'Modernization of the NPP with VVER reactor'. The following problems are considered: improving the NPP's safety and reliability; reactor modernization, the lifetime prolongation; increasing of the reactor operating characteristics; methods of capacity factor increasing: refueling control, maintenance control; technical and economical aspects of NPP modernization; modernization of the automated control system of the fuel process at the NPP's; technical features and methods for the continued radiation and technology control at the NPP's; training, increasing the staff qualification and NPP modernization

  18. Economic aspects of Dukovany NPP fuel cycle

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

  19. Neutron shielding calculation for VVER NPP

    There are two methods for neutron transport (shielding) calculation used in Energoproject, Prague, the method of discrete ordinates (code TORT-DORT) and the Monte Carlo method (codes MCNP and module within the code SCALE). The task concerning neutron dose rates calculation near casks with VVER spent fuel are presented as an example. Measured neutron dose rates of real loaded C-30 casks for VVER spent fuel assemblies are compared with calculated values in the frame of the international benchmark calculation task. A part of the task realized by the Atomic Energy Research (AER) organization concerning neutron shielding is calculated. The cask C-30 is used in Slovak Jaslovske Bohunice NPP for transport of spent fuel assemblies to the storage facility. The benchmark task has been calculated by the two-dimensional code DORT originated from Oak Ridge National Laboratory. The code solves transport problems using the method of discrete ordinates (SN - method). Calculated neutron dose rates in azimuth and vertical directions show good agreement with the experiment within the range of the measurement errors. In comparison with the other codes the results of DORT are approximately 20% lower. There have been analysed differences between one- and two- dimensional approach and influence of the flux-to-dose rate conversion factors set

  20. On-line maintenance at Cofrentes NPP

    Cofrentes NPP (CNPP) has developed a Level 1 PSA with the following scope: analysis of internal events, with the reactor initially operating at power; internal and external flooding risk analysis; internal fire risk analysis; reliability analysis of the containment heat removal and containment isolation systems. Level 1 CNPP-PSA results reveal that total core damage frequency in CNPP is less than other similar BWR/6 plants. The CNPP-PSA related activities and applications being carried out currently are: prioritization of motor operated valves related to GL-89/10; complementary analysis for exemption to some 10CFR50 App. J requirements; Q-List grading; risk-informed IST program; reliability-centered maintenance; maintenance rule support; on-line maintenance support; off-line risk-monitor development; PSA applicability to the 10CFR50 App. R requirements, analysis of the frequency of miss-oriented fuel bundle event, adjusting of MAAP 3.0B, revision 10, on VAX and PC; acquisition of MAAP 4; development of Level1/Level2-PSA interface; seismic site categorization for the IPEEE; etc. (author)

  1. Intranet portal at the Krsko NPP, Slovenia

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

  2. Lessons learnt from Ignalina NPP decommissioning project

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

  3. Tritium bearing molecular sieves from NPP Cernavoda

    Full text: Drying towers packed with molecular sieve beds are used to retain tritiated heavy water resulting in Cernavoda NPP current operation, which has leached from various parts of the reactor systems, in order to recover and reduce heavy water loses and to minimize tritium contamination. Molecular sieve during operation are put through cycles of adsorption and regeneration, in fact a desorption phase, and have a lifetime of several hundred cycles adsorption-desorption. After sending their life time the molecular sieves become tritium bearing radioactive wastes and have to be dealt with accordingly. The present paper will briefly describe the current practices for molecular sieve conditioning prior to disposal and the requirements for the conditioned radioactive wastes. Also, the paper presents the development of the conditioning technology for the tritiated molecular sieve capable of realizing a product which matches the waste acceptance requirements imposed by the National Authority for Control of Nuclear Activities, CNCAN, for the disposal at the DNDR Baita - Bihor national repository . (authors)

  4. Ultrasonic wavequide sensor for NPP acoustic monitoring

    Design of a waveguide sensor for NPP equipment acoustic testing is considered taking as an example a water coolant steam content monitor designed for application in an active emission-receipt regime. The sensor comprises an acoustic transducer, a waveguide with a suspension and a sensitive element. The transducer includes a disk piezoelement of TsTS-19 ceramics. A longitudinal wave waveguide, produced of a steel wire 0.8-1.2 mm in diameter, can transmit signals within the 50-1000 kHz range. A capillar tube 1.6x0.2 mm in diameter and 200 mm long with sealed ends is used as a sensitive element. The sensor operation is based on determining ultrasonic pulse attennuation in the capillar, which changes depending on acoustic wave resistance of the following-round coolant and depends on steam content. In passive regime the sensor may be applied for acoustic-emission monitoring of various equipment. In this case a matching device, providing for emission signal transmission from the monitored object surface to the waveguide, should be introduced instead of the sensitive element

  5. NPP Krsko simulator training for operations personnel

    Acquisition of a full scope replica simulator represents an important achievement for Nuclear power Plant Krsko. Operating nuclear power plant systems is definitely a set of demanding and complex tasks. The most important element in the goal of assuring capabilities for handling such tasks is efficient training of operations personnel who manipulate controls in the main control room. Use of a simulator during the training process is essential and can not be substituted by other techniques. This article gives an overview of NPP Krsko licensed personnel training historical background, current experience and plans for future training activities. Reactor operator initial training lasts approximately two and a half years. Training is divided into several phases, consisting of theoretical and practical segments, including simulator training. In the past, simulator initial training and annual simulator retraining was contracted, thus operators were trained on non-specific full scope simulators. Use of our own plant specific simulator and associated infrastructure will have a significant effect on the operations personnel training process and, in addition, will also support secondary uses, with the common goal to improve safe and reliable plant operation. A regular annual retraining program has successfully started. Use of the plant specific simulator assures consistent training and good management oversight, enhances conformity of operational practices and supports optimization of operating procedures. (author)

  6. LILW management in Paks NPP 2004

    Waste classification, sources, volumes, composition and treatment of different (liquid and solid) wastes in Paks are presented. The pretreatment of the liquid waste includes Collection in tanks, decanting, filtration (ion exchange resins), chemical treatment, evaporation, condensate cleaning and discharging, radioactive liquid wastes storage. Yearly accumulated volume of liquid wastes is approximately 250 m3; evaporation residues - 3644 m3 used ion exchange resins - 39.1 m3, and acidising solutions of evaporator - 190 m3. The free storage capacity for liquids is sufficient for 1-2 years, so 3800 m3 new capacities are under construction. The volume reduction technology used in Paks NPP is also presented in the paper. It is stated that the solidification will only be started after constructing the final disposal facility. The main steps of low and intermediate level solid wastes treatment are: selective collection, transport in the controlled area, pre-qualification (potentially exemptible or not), assorting, compaction by 50t compactor, wastes qualification, interim storage. The volume of solid wastes yearly is 600 - 1000 pcs. New elements in the concept for the low and intermediate level wastes are added after the fuel damage at unit 2 such as: additional interim storage capacities (3800 m3) for liquids (under construction); TRU separation (if needed); supercompaction technology and others because of new waste acceptance criteria under generation

  7. Training needs analysis at Cernavoda NPP - Romania

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

  8. SAT for NPP personnel training in Spain

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

  9. Regulation and license for NPP in Indonesia

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

  10. Network scheduling at Belene NPP construction site

    Four types of schedules differing in the level of their detail are singled out to enhance the efficiency of Belene NPP Project implementation planning and monitoring: Level 1 Schedule–Summary Integrated Overall Time Schedule (SIOTS) is an appendix to EPC Contract. The main purpose of SIOTS is the large scale presentation of the current information on the Project implementation. Level 2 Schedule–Integrated Overall Time Schedule (IOTS)is the contract schedule for the Contractor (ASE JSC) and their subcontractors.The principal purpose of IOTS is the work progress planning and monitoring, the analysis of the effect of activities implementation upon the progress of the Project as a whole. IOTS is the reporting schedule at the Employer –Contractor level. Level 3 Schedules, Detail Time Schedules(DTS) are developed by those who actually perform the work and are agreed upon with Atomstroyexport JSC.The main purpose of DTS is the detail planning of Atomstroyexport subcontractor's activities. DTSare the reporting schedules at the level of Contractor-Subcontractor. Level 4 Schedules are the High Detail Time Schedules (HDTS), which are the day-to-day plans of work implementation and are developed, as a rule, for a week's time period.Each lower level time schedule details the activities of the higher level time schedule

  11. NPP lifetime philosophy: the transatlantic difference

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

  12. Management of occupational exposure at Cernavoda NPP

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

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

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

  14. Implementing safety improvement program (SIP-3) at Ignalina NPP

    It is stated in license validity conditions for Ignalina NPP Unit 2 operation that the operating organization is to implement on time safety improvement measures and to submit an implementation report to VATESI on a quarterly basis. It is stated in another condition of validity of operation license for Ignalina NPP Unit 2 that based on the results of SAR-2 and RSR-2, the organization operating the Ignalina NPP is to prepare and submit to VATESI for approval the new safety improvement at Ignalina NPP program SIP-3 by December 2004. The Ignalina NPP submitted the draft SIP-3 in December 2004, and on April 8, 2005, VATESI approved it. One hundred and fifteen SIP-3 measures are to be implemented at Ignalina NPP in 2005-2008; 80, 24, and 7 measures in 2005, 2006, and 2007-2008, respectively, whereas 4 measures are being implemented continuously. In 2005, items were being implemented of SIP-3 that had not been accomplished when implementing SIP-2 in 2004 and envisaged in VATESI requirements and statements of inspections. The recommendations of safety case for the single Unit 2 operational at Ignalina NPP, those of SAR and RSR for INPP Unit 2, the plan of measures of INPP Unit 1 decommissioning program, the recommendations of the International Physical Protection Advisory Service (IPPAS) mission, instructions of INPP police team of the State Border Guard Service, technical decisions by INPP, statements, plans of measures, safety upgrading modifications, etc. were also taken into consideration when producing SIP-3. Major SIP-3 measures for safety upgrading at INPP implemented and approved in 2005 are listed. (author)

  15. Evolution of reactor monitoring and protection systems for PWR

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

  16. Corrosion product transfer in PWR primary circuits during cold shutdowns

    Two experimental tests have been performed to study the corrosion product transfer during PWR cold shutdowns: one with nickel ferrite and the other one with metallic nickel. The temperature evolution together with boron and oxygen concentration evolutions are similar to those obtained during cold shutdowns of PWR primary circuits. With metallic nickel, the increase of the Ni concentration occurs during the decrease of the primary temperature and mainly when the oxygenation is realised. Whereas, with nickel ferrite, the Ni concentration increase occurs during the 24 hours after the oxygenation. These results compared to the plant data lead to conclude that metallic nickel presence in the core is the most probable hypothesis. (author)

  17. The latest full-scale PWR simulator in Japan

    The latest MHI Full-scale Simulator has an excellent system configuration, in both flexibility and extendability, and has highly sophisticated performance in PWR simulation by the adoption of CANAC-II and PRETTY codes. It also has an instructive character to display the plant's internal status, such as RCS condition, through animation. Further, the simulation has been verified to meet a functional examination at model plant, and with a scale model test result in a two-phase flow event, after evaluation for its accuracy. Thus, the Simulator can be devoted to a sophisticated and broad training course on PWR operation. (author)

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

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

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

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

  20. Examination of dissimilar metal welds in BWR and PWR piping

    This paper addresses dissimilar metal weld examinations at PWRS. Surveys were conducted to document the dissimilar metal weld configurations at PWR plants and to update the information known about dissimilar metal weld configurations at BWR plants. The experiences which BWR utilities have had with dissimilar metal weld examinations are documented and include: correct identification of IGSCC, indications thought to be IGSCC but were actually fabrication flaws, and difficulties encountered with the examination of dissimilar metal welds after stress improvement. An experimental program was conducted which verified that the longitudinal wave procedures developed for BWRs are also applicable to PWR designs

  1. Evaluation of PWR and BWR pin cell benchmark results

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

  2. PHEDRE model for the simulation of PWR reactors

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE

  3. Approximation for maximum pressure calculation in containment of PWR reactors

    A correlation was developed to estimate the maximum pressure of dry containment of PWR following a Loss-of-Coolant Accident - LOCA. The expression proposed is a function of the total energy released to the containment by the primary circuit, of the free volume of the containment building and of the total surface are of the heat-conducting structures. The results show good agreement with those present in Final Safety Analysis Report - FSAR of several PWR's plants. The errors are in the order of ± 12%. (author)

  4. Post DNB heat transfer experiments for PWR fuel assemblies

    Nuclear Power Engineering Corporation (NUPEC) and Mitsubishi performed heat transfer experiments on post DNB (departure from nucleate boiling) for the pressurized water reactor (PWR) fuel assemblies under the sponsorship of the Japanese Ministry of Economy, Trade and Industry (METI) as one of a series of fuel assembly verification tests. Based on the obtained experimental data, a new evaluation model for the fuel rod heat transfer behavior after DNB was developed. A large safety margin, which had remained in the present thermal-hydraulic design that did not allow DNB, was confirmed by applying the developed model to the PWR plant safety analysis. (author)

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

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

    1997-04-01

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

  6. Advanced ion exchange resins for PWR condensate polishing

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

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

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

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

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

    2010-07-01

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

  9. Rod ejection accident 3D-dynamic analysis in Trillo NPP with RELAP5/PARCS V2.7 coupled codes

    The Rod Ejection Accident (REA) belongs to the Reactivity Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In previous works, we have analyzed this transient in Trillo NPP at different power levels at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5-MOD3.3/PARCSv2.7. In this work, we present the results of the REA analysis at 30% of the rated power at BOC. In the thermalhydraulic model used, each fuel assemblies has been modelled as an independent channel, for that, the RELAP5 source code has been modified and recompiled to accept this large number of channels. The neutronic nodal discretization consists of 177 x 32 active nodes, considering 28 different fuel elements with 867 neutronic compositions. The cross-sections sets are obtained from CASMO4-SIMULATE3 using the SIMTAB methodology developed in UPV. The transient departs from an initially critical core, being the withdrawal speed of the control rod a typical bounding value. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods. (authors)

  10. Rod ejection accident 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS V2.7 coupled codes

    The Rod Ejection Accident (REA) belongs to the Reactivity Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In previous works, we have analyzed this transient in Almaraz NPP at different power levels at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5-MOD3.3/PARCS v2.7. In this work, we present the results of the REA analysis at hot zero power at BOC with all control rods inserted. In the thermal-hydraulic model used, each fuel assemblies has been modelled as an independent channel, for that, the RELAP5 source code has been modified and recompiled to accept this large number of channels. The neutronic nodal discretization consists of 157 x 24 active nodes, considering 13 different fuel elements with 291 neutronic compositions. The cross-sections sets are obtained from CASMO4-SIMULATE3 using the SIMTAB methodology developed in UPV. The transient departs from an initially critical core, being the withdrawal speed of the control rod a typical bounding value. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods. (authors)

  11. A comparison of the Kozloduy NPP and Temelin NPP I and C projects

    Because Kozloduy NPP and Temelin NPP are both VVER 1000 plants of roughly the same vintage, they are very similar in design. However, from the viewpoint of their I and C modernization projects, there are significant differences between these plants. Some of these differences stem from the evolution of I and C technology over the relatively short period between the two projects. Other differences arise from the fact that the Kozloduy project is a phased upgrade of the I and C systems in an operating plant while the Temelin project was a 'one time' installation of the entire plant I and C system. This paper discusses these differences as well as trends in the nuclear I and C field that will shape the industry in the future. In addition to technology evolution, the comparative advantages or problems in phased upgrade versus 'one time' installations are discussed. Conclusions drawn provide insight for the planning of future I and C upgrades in VVERs and other types of nuclear power plants. (author)

  12. Equipment reliability improvement process; implementation in Almaraz NPP and Trillo NPP

    The Equipment Reliability Improvement Process (INPO AP-913) is a non-regulatory process developed by the US Nuclear Industry for improving Plants Availability. This Process integrates and coordinates a broad range of equipment reliability activities into one process, performed by the Plant in a non-centralized way. The integration and coordination of these activities will allow plant personnel to evaluate the trends of important station equipment, develop and implement long-term equipment health plans, monitor equipment performance and condition, and make adjustments to preventive maintenance tasks and frequencies based on equipment operating experience, if necessary, arbitrating operational and design improvements, to reach a Failure-free Operation. This paper describes the methodology of Equipment Reliability Improvement Process, being focused on main aspects of the implementation process, relating to the scope and establishment of an Equipment Reliability Monitoring Plan, which should include and complement the existing mechanisms and organizations in the Plant to monitor the condition and performance of the equipments, with the common aim of achieving an operation free of failures. The paper will describe the tools that Iberdrola Ingenieria has developed to support the implementation and monitoring of the Equipment Reliability Improvement Process, as well as the results and lessons learned from its implementation in Almaraz NPP and Trillo NPP. (authors)

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

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

  14. Technical innovations at NPP Dukovany - for safe and efficient operation

    Inherent features of the NPP Dukovany design provide a significant confidence in its nuclear safety assurance; among these features should be emphasised the reactor core stability and its control and protection system capability to hold the reactor in safe state following scram or accident conditions. Nevertheless, NPP Dukovany was designed in the early seventies, and current requirements for nuclear safety assurance are more strict and/or specific as a result of the technical development and lessons learned from nuclear accidents. The paper compares the safety design base established at the time of NPP Dukovany project implementation and the current reference design base. The paper also presents procedures applied to implement technical and operational measures which are introduced to fulfil the current basic safety criteria. The scope of such measures applied at NPP Dukovany is compared with that of EU countries introduced for the same reason - to meet the updated safety related requirements. Examples of some innovations already implemented or under implementation give an idea how NPP Dukovany proceeds in reaching the goal of harmonising its safety with the requirements to be met before the Czech Republic becomes a member country of the European Union. (author)

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

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

  16. Spent fuel pool risk analysis for the Dukovany NPP

    UJV Rez, a.s. maintains a Living Probabilistic Safety Assessment (Living PSA) program for Dukovany Nuclear Power Plant (NPP) in the Czech Republic. This project has been established as a framework for activities related to risk assessment and to support for risk-informed decision making at this plant. The most extensively used PSA application at Dukovany NPP is risk monitoring of instantaneous (point-in-time) risk during plant operation, especially for the purpose of configuration risk management during plant scheduled outages to avoid risk significant configurations. The scope of PSA for Dukovany NPP includes also determination of a risk contribution from spent fuel pool (SFP) operation to provide recommendations for the prevention and mitigation of SFP accidents and to be applicable for configuration risk management. This paper describes the analysis of internal initiating events (IEs) in PSA for Dukovany NPP, which can contribute to the risk from SFP operation. The analysis of those IEs was done more thoroughly in the PSA for Dukovany NPP in order to be used in instantaneous risk monitoring. (orig.)

  17. Methodology for the LABIHS PWR simulator modernization

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

    2011-07-01

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

  18. Methodology for the LABIHS PWR simulator modernization

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

  19. Crack growth rate of PWR piping

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 2800C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 2800C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

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

    Lind, Morten; Zhang, Xinxin

    2014-01-01

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